ML23304A358

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LLC Revision 1 to Standard Design Approval Application, Part 2, Chapter 11, Radioactive Waste Management
ML23304A358
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ale US460 Plant ard Design Approval Application pter Eleven ioactive Waste nagement Safety Analysis Report n 1 uScale Power LLC. All Rights Reserved

document bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy of the information in this document, other than by the U.S. Nuclear Regulatory Commission C), is authorized without the express, written permission of NuScale Power, LLC.

NRC is permitted to make the number of copies of the information contained in these reports ded for its internal use in connection with generic and plant-specific reviews and approvals, well as the issuance, denial, amendment, transfer, renewal, modification, suspension, ocation, or violation of a license, permit, order, or regulation subject to the requirements of CFR 2.390 regarding restrictions on public disclosure to the extent such information has been tified as proprietary by NuScale Power, LLC, copyright protection notwithstanding.

arding nonproprietary versions of these reports, the NRC is permitted to make the number of itional copies necessary to provide copies for public viewing in appropriate docket files in lic document rooms in Washington, DC, and elsewhere as may be required by NRC ulations. Copies made by the NRC must include this copyright notice in all instances and the prietary notice if the original was identified as proprietary.

APTER 11 RADIOACTIVE WASTE MANAGEMENT . . . . . . . . . . . . . . . . . . . . 11.1-1 11.1 Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 11.1.1 Design Basis Reactor Coolant Activity . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 11.1.2 Design Basis Secondary Coolant Activity . . . . . . . . . . . . . . . . . . . . . . . 11.1-3 11.1.3 Realistic Reactor Coolant and Secondary Coolant Activity Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-3 11.1.4 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-4 11.2 Liquid Waste Management System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.3 Radioactive Effluent Releases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-5 11.2.4 Testing and Inspection Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-6 11.2.5 Instrumentation and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-6 11.2.6 Reference. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-7 11.3 Gaseous Waste Management System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-1 11.3.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-1 11.3.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-1 11.3.3 Radioactive Effluent Releases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-5 11.3.4 Ventilation Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-6 11.3.5 Instrumentation and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-6 11.3.6 Reference. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-7 11.4 Solid Waste Management System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-1 11.4.1 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-1 11.4.2 Radioactive Effluent Releases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-5 11.4.3 Malfunction Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-5 11.4.4 Testing and Inspection Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-5 11.4.5 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-5 11.5 Process and Effluent Radiation Monitoring Instrumentation and Sampling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-1 11.5.1 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-1 11.5.2 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-3 cale US460 SDAA i Revision 1

11.6 Instrumentation and Control Design Features for Process and Effluent Radiological Monitoring, and Area Radiation and Airborne Radioactivity Monitoring. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.6-1 cale US460 SDAA ii Revision 1

le 11.1-1: Maximum Core Isotopic Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-5 le 11.1-2: Parameters Used to Calculate Coolant Source Terms . . . . . . . . . . . . . . . 11.1-6 le 11.1-3: Specific Parameters for CRUD . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-7 le 11.1-4: Primary Coolant Design Basis Source Term . . . . . . . . . . . . . . . . . . . . . . . 11.1-8 le 11.1-5: Secondary Coolant Design Basis Source Term. . . . . . . . . . . . . . . . . . . . 11.1-10 le 11.1-6: Primary Coolant Realistic Source Term. . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-12 le 11.1-7: Secondary Coolant Realistic Source Term . . . . . . . . . . . . . . . . . . . . . . . 11.1-14 le 11.1-8: Tritium Concentration versus Primary Coolant Recycling Modes . . . . . . 11.1-16 le 11.2-1: Major Component Design Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-8 le 11.2-2: Off-Normal Operation and Anticipated Operational Occurrence Consequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-9 le 11.2-3: Expected Liquid Waste Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-10 le 11.2-4: Liquid Effluent Release Calculation Inputs . . . . . . . . . . . . . . . . . . . . . . . 11.2-11 le 11.2-5: Estimated Annual Releases to Liquid Radioactive Waste System Discharge Header . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-12 le 11.2-6: LADTAP II Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-14 le 11.2-7: Liquid Effluent Dose Results for 10 CFR 50 Appendix I . . . . . . . . . . . . . 11.2-15 le 11.2-8: Classification of Structures, Systems, and Components . . . . . . . . . . . . . 11.2-16 le 11.2-9: Liquid Release Concentrations Compared to 10 CFR 20 Appendix B Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-17 le 11.3-1: Gaseous Radioactive Waste System Design Parameters. . . . . . . . . . . . . 11.3-8 le 11.3-2: Major Equipment Design Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-9 le 11.3-3: Gaseous Radioactive Waste System Equipment Malfunction Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-10 le 11.3-4: Gaseous Effluent Release Calculation Inputs . . . . . . . . . . . . . . . . . . . . . 11.3-12 le 11.3-5: Gaseous Estimated Discharge for Normal Effluents . . . . . . . . . . . . . . . . 11.3-13 le 11.3-6: GASPAR Code Input Parameter Values . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-17 le 11.3-7: Gaseous Effluent Dose Results for 10 CFR 50 Appendix I . . . . . . . . . . . 11.3-18 le 11.3-8: Gaseous Effluent Dose Evaluation for Gaseous Radioactive Waste System Failure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-19 le 11.3-9: Vapor Condenser Package Assembly Radiological Content . . . . . . . . . . 11.3-20 le 11.3-10: Classification of Structures, Systems, and Components . . . . . . . . . . . . . 11.3-21 le 11.4-1: List of Systems, Structures, and Components Design Parameters . . . . . . 11.4-7 le 11.4-2: Estimated Annual Volumes of Dry Solid Waste . . . . . . . . . . . . . . . . . . . . . 11.4-8 cale US460 SDAA iii Revision 1

le 11.4-3: Estimated Annual Volumes of Wet Solid Waste . . . . . . . . . . . . . . . . . . . . 11.4-9 le 11.4-4: Solid Radioactive Waste System Equipment Malfunction Analysis . . . . . 11.4-10 le 11.4-5: Classification of Structures, Systems, and Components . . . . . . . . . . . . . 11.4-12 le 11.5-1: Process and Effluent Radiation Monitoring Instrumentation Characteristics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-4 le 11.5-2: Provisions for Sampling Gaseous Process and Effluent Streams . . . . . . . 11.5-9 le 11.5-3: Provisions for Sampling Liquid Process and Effluent Streams . . . . . . . . 11.5-10 le 11.5-4: Effluent and Process Monitoring Off Normal Radiation Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-11 cale US460 SDAA iv Revision 1

ure 11.2-1a: Liquid Radioactive Waste System Diagram. . . . . . . . . . . . . . . . . . . . . . . 11.2-19 ure 11.2-1b: Liquid Radioactive Waste System Diagram. . . . . . . . . . . . . . . . . . . . . . . 11.2-20 ure 11.2-1c: Liquid Radioactive Waste System Diagram. . . . . . . . . . . . . . . . . . . . . . . 11.2-21 ure 11.2-1d: Liquid Radioactive Waste System Diagram. . . . . . . . . . . . . . . . . . . . . . . 11.2-22 ure 11.2-1e: Liquid Radioactive Waste System Diagram. . . . . . . . . . . . . . . . . . . . . . . 11.2-23 ure 11.2-1f: Liquid Radioactive Waste System Diagram. . . . . . . . . . . . . . . . . . . . . . . 11.2-24 ure 11.2-1g: Liquid Radioactive Waste System Diagram. . . . . . . . . . . . . . . . . . . . . . . 11.2-25 ure 11.2-1h: Liquid Radioactive Waste System Diagram. . . . . . . . . . . . . . . . . . . . . . . 11.2-26 ure 11.2-1i: Liquid Radioactive Waste System Diagram. . . . . . . . . . . . . . . . . . . . . . . 11.2-27 ure 11.2-1j: Liquid Radioactive Waste System Diagram. . . . . . . . . . . . . . . . . . . . . . . 11.2-28 ure 11.3-1: Gaseous Radioactive Waste System Diagram . . . . . . . . . . . . . . . . . . . . 11.3-22 ure 11.4-1: Block Diagram of the Solid Radioactive Waste System. . . . . . . . . . . . . . 11.4-13 ure 11.4-2a: Process Flow Diagram for Wet Solid Waste . . . . . . . . . . . . . . . . . . . . . . 11.4-14 ure 11.4-2b: Solid Radioactive Waste System Diagram . . . . . . . . . . . . . . . . . . . . . . . 11.4-15 ure 11.5-1a: Radioactive Effluent Flow Paths with Process and Effluent Radiation Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-14 ure 11.5-1b: Radioactive Effluent Flow Paths with Process and Effluent Radiation Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-15 ure 11.5-2: Process and Effluent Radiation Monitoring System Instrumentation and Control Configuration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-16 ure 11.5-3: Off-Line Radiation Monitor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-17 ure 11.5-4: Adjacent-to-Line Radiation Monitor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-18 ure 11.5-5: In-Line Radiation Monitor. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-19 ure 11.5-6: Reactor Building HVAC System Plant Exhaust Stack Effluent Radiation Monitor. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-20 cale US460 SDAA v Revision 1

1 Source Terms Sources of radioactivity in the primary coolant are created by fission and activation processes in the reactor. The NuScale Power Plant US460 standard design uses a shared systems source term for systems that receive input from all six modules, such as radwaste systems. The shared systems source term is calculated as one module running at a design basis failed fuel fraction and five modules running at a realistic failed fuel fraction. The secondary coolant may become contaminated by primary-to-secondary leakage through the steam generator. This section discusses two source terms for the primary and secondary coolants: a design basis source term and a realistic source term.

The design basis source term provides a basis for design capacities of waste management components, performance of the waste management systems and design of radiological monitoring equipment. The design basis source term is also used for the evaluation of shielding (General Design Criterion 61). The coolant source terms used for dose consequences of design basis events are found in Section 15.0. Equipment qualification is discussed in Section 3.11.

A realistic source term is used to calculate the quantity of radioactive materials released annually in liquid and gaseous effluents during normal plant operations, including anticipated operational occurrences, to demonstrate compliance with effluent limits of 10 CFR 20 Appendix B, Table 2, and the as low as reasonably achievable objectives of 10 CFR 50 Appendix I. The methodology used to develop the primary and secondary coolant realistic source terms is described in TR-123242, (Reference 11.1-1).

The plant is designed with up to six NuScale Power Modules (NPMs) partially immersed in a pool of water, called the reactor pool. Because of this design, there is a potential for neutron activation of the reactor pool water. Additionally, given the relative proximity of the secondary coolant to the reactor core, there is the potential for neutron activation of the secondary coolant. However, the production of radionuclides in the secondary coolant is several orders of magnitude less than that in the primary coolant and is considered negligible. The production of radionuclides in the reactor pool water is discussed in Section 12.2.1.

1.1 Design Basis Reactor Coolant Activity The design basis source term uses the same methodology described in TR-123242, except it assumes that fuel defects are an order-of-magnitude greater than the realistic coolant source term (Table 11.1-2). These defects are assumed to be uniformly distributed throughout the reactor core. The primary coolant design basis source term is provided in Table 11.1-4.

1.1.1 Fission Products The isotopic inventory is developed for a single fuel assembly irradiated to the value presented in Table A-1 in TR-123242 (Reference 11.1-1). The quantity of cale US460 SDAA 11.1-1 Revision 1

The parameters used in the calculation of the coolant source terms, including values for the fission product escape rate coefficients, coolant cleanup rate, and demineralizer effectiveness, are listed in Table 11.1-2. The quantity of fission products in the fuel pins and the release to the primary coolant are calculated using the methodology in TR-123242 (Reference 11.1-1).

1.1.2 Activation Products The permeation of tritium from the fuel to the primary coolant is modeled using the Electric Power Research Institute (EPRI) Tritium Management Model (Reference 11.1-3). The tritium permeation rate is linearly scaled from 4100 MWt to 250 MWt, and adjusted for a 95 percent capacity factor, as shown in Table 11.1-2.

In the primary coolant system, neutron activation of various constituents in the water forms activation products. These activation products are independent of the failed fuel fraction. The neutron activation products include N-16, H-3, Ar-41, and C-14.

Because of its short half-life, N-16 is not of concern for offsite dose considerations. Table 12.2-4 lists the N-16 concentration at various locations in the primary coolant loop.

The predominant tritium production reactions are high-energy neutron interactions with lithium and boron isotopes.

The concentrations of tritium in the coolant streams vary depending on whether the primary coolant letdown to liquid radioactive waste system is recycled to the reactor pool, recycled back to chemical and volume control system (CVCS) makeup, or discharged through liquid radioactive waste system. The various tritium concentrations are presented in Table 11.1-8.

In the absence of significant N-16 in the primary coolant near the steam generators, natural argon can be injected into the primary coolant to improve the sensitivity of primary-to-secondary leak rate calculations. When the injected argon is activated in the reactor core, Ar-41 is produced, which can be used to detect leakage from the primary system.

The C-14 primary coolant equilibrium activity is calculated using the following equation:

A C14 = P C14 x f Eq. 11.1-1 where, AC14 = The equilibrium activity of C-14 in the primary coolant (Ci),

cale US460 SDAA 11.1-2 Revision 1

f = The fraction of C-14 retained in the primary coolant.

1.1.3 Corrosion Products The concentration values of radioactive corrosion and wear products in the reactor coolant are developed using guidance from ANSI/ANS 18.1-1999 (Reference 11.1-2). The specific parameters for adjusting the ANSI/ANS 18.1-1999 reference values are listed in Table 11.1-3.

1.2 Design Basis Secondary Coolant Activity The design basis secondary coolant activity is determined from an assumed primary-to-secondary leak rate (Table 11.1-2) assuming a design basis primary coolant activity concentration (Table 11.1-4). The secondary coolant design basis source term is listed in Table 11.1-5.

1.2.1 Steam Generator Leakage For the radionuclides that enter the secondary coolant, various removal mechanisms are also incorporated that affect the equilibrium concentration in the secondary coolant. The removal mechanisms include steam leaks in the Turbine Building, condensate polishers, and radioactive decay.

1.2.2 Noble Gases in Secondary Coolant Activity Source Term Noble gases are removed in the secondary coolant by the condenser air removal system. Therefore, only pass-through concentrations of noble gases are assumed to be present in the steam generators. The concentration of noble gases in the secondary coolant is calculated by multiplying the concentration of the noble gas in the primary coolant by the primary-to-secondary leak rate, and dividing by the sum of the secondary flow rate and primary-to-secondary leak rate. The secondary coolant noble gas concentration after passing through the condenser is negligible.

1.2.3 Other Isotopes in Secondary Coolant Activity Source Term For radioisotopes other than tritium, it is assumed there are no steam leaks. The modeling process for secondary coolant concentrations is described in Section 4.2 of Reference 11.1-1. The condensate system is designed such that 100 percent of the secondary coolant flow passes through the condensate demineralizers. The parameters used to calculate the secondary design basis source term are listed in Table 11.1-2 and Table 11.1-4.

1.3 Realistic Reactor Coolant and Secondary Coolant Activity Source Terms A realistic source term is used to evaluate normal expected effluent releases as described in Section 11.2 and Section 11.3.

cale US460 SDAA 11.1-3 Revision 1

Table 11.1-7.

Details of the release modeling are presented in Section 11.2 and Section 11.3.

The resultant airborne concentrations are presented in Section 12.2.

1.4 References 11.1-1 NuScale Power, LLC, Effluent Release (GALE Replacement)

Methodology and Results, TR-123242-P, Revision 1.

11.1-2 American National Standards Institute/American Nuclear Society, "Radioactive Source Term for Normal Operation of Light Water Reactors,"

ANSI/ANS 18.1-1999, LaGrange Park, IL.

11.1-3 Electric Power Research Institute, Inc., "EPRI Tritium Management Model," EPRI #1009903, EPRI, Palo Alto, CA, 2005.

cale US460 SDAA 11.1-4 Revision 1

Nuclide Core Inventory (Ci) Nuclide Core Inventory (Ci)

Noble Gases Other Fission Products Kr83m 7.3E+05 Y91 7.0E+06 Kr85m 1.5E+06 Y92 7.6E+06 Kr85 1.5E+05 Y93 8.9E+06 Kr87 2.8E+06 Zr97 1.1E+07 Kr88 3.7E+06 Nb95 1.1E+07 Kr89 4.6E+06 Mo99 1.3E+07 Xe131m 1.0E+05 Mo101 1.2E+07 Xe133m 4.5E+05 Tc99m 1.1E+07 Xe133 1.4E+07 Tc99 2.2E+02 Xe135m 3.3E+06 Ru103 1.4E+07 Xe135 4.1E+06 Ru105 1.1E+07 Xe137 1.2E+07 Ru106 8.6E+06 Xe138 1.1E+07 Rh103m 1.4E+07 Halogens Rh105 1.0E+07 Br82 4.0E+04 Rh106 9.6E+06 Br83 7.1E+05 Ag110 3.6E+06 Br84 1.2E+06 Sb124 2.1E+04 Br85 1.5E+06 Sb125 1.5E+05 I129 5.5E-01 Sb127 8.3E+05 I130 4.2E+05 Sb129 2.4E+06 I131 7.2E+06 Te125m 3.6E+04 I132 1.0E+07 Te127m 1.3E+05 I133 1.4E+07 Te127 8.2E+05 I134 1.5E+07 Te129m 3.9E+05 I135 1.3E+07 Te129 2.3E+06 Rubidium, Cesium Te131m 1.5E+06 Rb86m 3.1E+03 Te131 6.1E+06 Rb86 2.5E+04 Te132 1.0E+07 Rb88 3.8E+06 Te133m 6.4E+06 Rb89 5.0E+06 Te134 1.2E+07 Cs132 5.2E+02 Ba137m 1.7E+06 Cs134 3.4E+06 Ba139 1.2E+07 Cs135m 4.4E+04 Ba140 1.2E+07 Cs136 7.8E+05 La140 1.2E+07 Cs137 1.8E+06 La141 1.1E+07 Cs138 1.3E+07 La142 1.0E+07 Other Fission Products Ce141 1.1E+07 P32 1.1E+03 Ce143 9.9E+06 Co57 8.1E+00 Ce144 9.2E+06 Sr89 5.2E+06 Pr143 9.6E+06 Sr90 1.2E+06 Pr144 9.3E+06 Sr91 6.8E+06 Np239 1.9E+08 Sr92 7.5E+06 C14 2.2E+01 Y90 1.2E+06 H3 2.1E+04 Y91m 4.0E+06 cale US460 SDAA 11.1-5 Revision 1

Parameter Value ctor core thermal power 250 + 5 = 255 MWt (102%)

ber of fuel assemblies in one core 37 ge of U-235 fuel enrichment 1.5% - 4.95%

nium mass in one fuel assembly 250.6 kg imum fuel assembly burnup 62,GWD/MTU ed fuel fractions:

PM Realistic source term 0.0066%

PM Design basis source term 0.066%

ape rate coefficients:

e, Kr gases 6.5E-08 s-1 Br, Cs, Rb 1.3E-08 s-1 o, Tc, Ag 2.0E-09 s-1 e 1.0E-09 s-1 r, Ba 1.0E-11 s-1 thers 1.6E-12 s-1 rage density of reactor coolant 0.71 gram/cm3 mass 1.0E+05 lb n injection concentration:

esign basis 0.15 µCi/cm3 ealistic 0.10 µCi/cm3 S flow rate (purification) 180 lb/min ondary coolant mass 5.0E+04 lb ondary steam leak rate:

esign Basis 1700 lb/hr/NPM ealistic 125 lb/hr/NPM ondary coolant flow rate 6.5E+05 lb/hr ontamination factors for CVCS mixed bed demineralizers:

alogens 100 s, Rb 2 ther 50 ontamination factors for condensate demineralizers:

alogens 100 s, Rb 10 ther 100 ary-to-secondary leak rate:

esign basis 75 lb/day/NPM ealistic 5.5 lb/day/NPM um permeation rate 9 Ci/yr bon-14 primary coolant retention rate 0.01 bon-14 removal rate by CVCS demineralizers 0 cale US460 SDAA 11.1-6 Revision 1

Parameter Symbol Units Value rmal power P MWt 250 ght of water in RCS WP kg 4.7E+04 own flow rate for purification FD kg/s 1.4 own flow rate for boron control FB kg/s 3.9E-03 through cation demineralizer FA kg/s 0 tion of material removed by cation demineralizer NA --- 0.9 tion of material removed by purification demineralizer NB --- 0.98 cale US460 SDAA 11.1-7 Revision 1

Primary Coolant Primary Coolant Nuclide Concentrations Nuclide Concentrations (Ci/g) (Ci/g)

Noble Gases Other FPs (continued)

Kr83m 7.7E-09 Mo101 4.3E-10 Kr85m 3.2E-08 Tc99m 1.1E-08 Kr85 5.9E-06 Tc99 2.1E-13 Kr87 1.8E-08 Ru103 1.1E-11 Kr88 5.1E-08 Ru105 3.6E-12 Kr89 1.2E-09 Ru106 6.8E-12 Xe131m 1.4E-07 Rh103m 1.1E-11 Xe133m 1.2E-07 Rh105 7.4E-12 Xe133 8.6E-06 Rh106 6.8E-12 Xe135m 1.1E-08 Ag110 8.1E-12 Xe135 2.3E-07 Sb124 1.6E-14 Xe137 3.9E-09 Sb125 1.2E-13 Xe138 1.3E-08 Sb127 6.1E-13 Halogens Sb129 7.6E-13 Br82 2.1E-10 Te125m 1.8E-11 Br83 1.2E-09 Te127m 6.7E-11 Br84 5.7E-10 Te127 2.6E-10 Br85 6.9E-11 Te129m 1.9E-10 I129 3.5E-15 Te129 2.7E-10 I130 1.7E-09 Te131m 6.2E-10 I131 4.4E-08 Te131 3.1E-10 I132 2.1E-08 Te132 4.6E-09 I133 6.7E-08 Te133m 3.9E-10 I134 1.2E-08 Te134 5.6E-10 I135 4.3E-08 Ba137m 2.1E-08 Rubidium, Cesium Ba139 1.0E-11 Rb86m 5.2E-14 Ba140 5.6E-11 Rb86 3.0E-10 La140 1.6E-11 Rb88 5.1E-08 La141 3.2E-12 Rb89 2.4E-09 La142 1.5E-12 Cs132 6.0E-12 Ce141 8.6E-12 Cs134 4.3E-08 Ce143 6.5E-12 Cs135m 3.6E-11 Ce144 7.3E-12 Cs136 9.4E-09 Pr143 7.6E-12 Cs137 2.2E-08 Pr144 7.2E-12 Cs138 1.9E-08 Np239 1.4E-10 Other FPs Corrosion/Activation Products - CRUD P32 8.4E-16 Na24 1.4E-08 Co57 6.4E-18 Cr51 7.7E-10 Sr89 3.8E-11 Mn54 4.0E-10 Sr90 6.0E-12 Fe55 3.0E-10 Sr91 2.0E-11 Fe59 7.5E-11 Sr92 1.1E-11 Co58 1.1E-09 Y90 1.5E-12 Co60 1.3E-10 Y91m 1.1E-11 Ni63 6.6E-11 Y91 5.6E-12 Zn65 1.3E-10 Y92 9.1E-12 Zr95 9.7E-11 Y93 4.3E-12 Ag110m 3.2E-10 cale US460 SDAA 11.1-8 Revision 1

Primary Coolant Primary Coolant Nuclide Concentrations Nuclide Concentrations (Ci/g) (Ci/g)

Zr97 6.3E-12 W187 7.0E-10 Nb95 9.1E-12 Water Activation Products Mo99 1.1E-08 C14 2.6E-10 Ar41 2.1E-07 cale US460 SDAA 11.1-9 Revision 1

Secondary Coolant Secondary Coolant Nuclide Concentrations Nuclide Concentrations (Ci/g) (Ci/g)

Noble Gases Other FPs (continued)

Kr83m 3.7E-14 Mo101 1.7E-15 Kr85m 1.6E-13 Tc99m 5.1E-14 Kr85 2.8E-11 Tc99 1.0E-18 Kr87 8.5E-14 Ru103 5.3E-17 Kr88 2.5E-13 Ru105 1.7E-17 Kr89 5.6E-15 Ru106 3.3E-17 Xe131m 6.6E-13 Rh103m 4.9E-17 Xe133m 5.6E-13 Rh105 3.6E-17 Xe133 4.2E-11 Rh106 4.4E-18 Xe135m 5.3E-14 Ag110 4.4E-18 Xe135 1.1E-12 Sb124 7.9E-20 Xe137 1.9E-14 Sb125 5.9E-19 Xe138 6.3E-14 Sb127 3.0E-18 Halogens Sb129 3.6E-18 Br82 1.0E-15 Te125m 8.5E-17 Br83 5.8E-15 Te127m 3.2E-16 Br84 2.5E-15 Te127 1.3E-15 Br85 1.6E-16 Te129m 9.3E-16 I129 1.7E-20 Te129 1.3E-15 I130 8.4E-15 Te131m 3.0E-15 I131 2.2E-13 Te131 1.3E-15 I132 9.8E-14 Te132 2.2E-14 I133 3.3E-13 Te133m 1.8E-15 I134 5.5E-14 Te134 2.5E-15 I135 2.1E-13 Ba137m 4.5E-14 Rubidium, Cesium Ba139 4.8E-17 Rb86m 6.1E-20 Ba140 2.7E-16 Rb86 1.6E-15 La140 7.9E-17 Rb88 2.3E-13 La141 1.5E-17 Rb89 1.0E-14 La142 7.2E-18 Cs132 3.2E-17 Ce141 4.2E-17 Cs134 2.3E-13 Ce143 3.2E-17 Cs135m 1.8E-16 Ce144 3.5E-17 Cs136 5.0E-14 Pr143 3.7E-17 Cs137 1.2E-13 Pr144 3.0E-17 Cs138 9.3E-14 Np239 6.6E-16 Other FPs Corrosion/Activation Products - CRUD P32 4.1E-21 Na24 6.7E-14 Co57 3.1E-23 Cr51 3.8E-15 Sr89 1.9E-16 Mn54 1.9E-15 Sr90 2.9E-17 Fe55 1.4E-15 Sr91 9.6E-17 Fe59 3.6E-16 Sr92 5.1E-17 Co58 5.6E-15 Y90 7.0E-18 Co60 6.4E-16 Y91m 4.9E-17 Ni63 3.2E-16 Y91 2.7E-17 Zn65 6.1E-16 Y92 4.3E-17 Zr95 4.7E-16 Y93 2.1E-17 Ag110m 1.6E-15 cale US460 SDAA 11.1-10 Revision 1

Secondary Coolant Secondary Coolant Nuclide Concentrations Nuclide Concentrations (Ci/g) (Ci/g)

Zr97 3.1E-17 W187 3.4E-15 Nb95 4.4E-17 Water Activation Products Mo99 5.5E-14 C14 1.3E-15 Ar41 1.0E-12 cale US460 SDAA 11.1-11 Revision 1

Primary Coolant Primary Coolant Nuclide Concentrations Nuclide Concentrations (Ci/g) (Ci/g)

Noble Gases Other FPs (continued)

Kr83m 7.7E-10 Mo101 4.3E-11 Kr85m 3.2E-09 Tc99m 1.0E-09 Kr85 1.6E-07 Tc99 2.1E-14 Kr87 1.8E-09 Ru103 1.1E-12 Kr88 5.1E-09 Ru105 3.6E-13 Kr89 1.2E-10 Ru106 6.8E-13 Xe131m 1.3E-08 Rh103m 1.1E-12 Xe133m 1.1E-08 Rh105 7.3E-13 Xe133 8.3E-07 Rh106 6.8E-13 Xe135m 1.1E-09 Ag110 4.8E-12 Xe135 2.3E-08 Sb124 1.6E-15 Xe137 3.9E-10 Sb125 1.2E-14 Xe138 1.3E-09 Sb127 6.1E-14 Halogens Sb129 7.6E-14 Br82 2.1E-11 Te125m 1.7E-12 Br83 1.2E-10 Te127m 6.6E-12 Br84 5.7E-11 Te127 2.6E-11 Br85 6.9E-12 Te129m 1.9E-11 I129 3.5E-16 Te129 2.7E-11 I130 1.7E-10 Te131m 6.2E-11 I131 4.4E-09 Te131 3.1E-11 I132 2.1E-09 Te132 4.6E-10 I133 6.7E-09 Te133m 3.9E-11 I134 1.2E-09 Te134 5.6E-11 I135 4.3E-09 Ba137m 2.1E-09 Rubidium, Cesium Ba139 1.0E-12 Rb86m 5.2E-15 Ba140 5.6E-12 Rb86 3.0E-11 La140 1.6E-12 Rb88 5.1E-09 La141 3.2E-13 Rb89 2.4E-10 La142 1.5E-13 Cs132 6.0E-13 Ce141 8.6E-13 Cs134 4.3E-09 Ce143 6.5E-13 Cs135m 3.6E-12 Ce144 7.3E-13 Cs136 9.4E-10 Pr143 7.6E-13 Cs137 2.2E-09 Pr144 7.2E-13 Cs138 1.9E-09 Np239 1.4E-11 Other FPs Corrosion/Activation Products - CRUD P32 8.4E-17 Na24 1.4E-08 Co57 6.4E-19 Cr51 7.7E-10 Sr89 3.8E-12 Mn54 4.0E-10 Sr90 5.9E-13 Fe55 3.0E-10 Sr91 2.0E-12 Fe59 7.5E-11 Sr92 1.1E-12 Co58 1.1E-09 Y90 1.4E-13 Co60 1.3E-10 Y91m 1.1E-12 Ni63 6.6E-11 Y91 5.6E-13 Zn65 1.3E-10 Y92 9.0E-13 Zr95 9.7E-11 Y93 4.3E-13 Ag110m 3.2E-10 cale US460 SDAA 11.1-12 Revision 1

Primary Coolant Primary Coolant Nuclide Concentrations Nuclide Concentrations (Ci/g) (Ci/g)

Zr97 6.3E-13 W187 7.0E-10 Nb95 1.6E-12 Water Activation Products Mo99 1.1E-09 C14 2.6E-10 Ar41 1.4E-07 cale US460 SDAA 11.1-13 Revision 1

Secondary Coolant Secondary Coolant Nuclide Concentrations Nuclide Concentrations (Ci/g) (Ci/g)

Noble Gases Other FPs (continued)

Kr83m 2.7E-16 Mo101 1.3E-17 Kr85m 1.1E-15 Tc99m 3.7E-16 Kr85 5.7E-14 Tc99 7.6E-21 Kr87 6.2E-16 Ru103 3.9E-19 Kr88 1.8E-15 Ru105 1.3E-19 Kr89 4.2E-17 Ru106 2.4E-19 Xe131m 4.5E-15 Rh103m 3.6E-19 Xe133m 4.0E-15 Rh105 2.6E-19 Xe133 2.9E-13 Rh106 3.2E-20 Xe135m 3.9E-16 Ag110 1.9E-19 Xe135 8.1E-15 Sb124 5.8E-22 Xe137 1.4E-16 Sb125 4.3E-21 Xe138 4.7E-16 Sb127 2.2E-20 Halogens Sb129 2.7E-20 Br82 7.6E-18 Te125m 6.2E-19 Br83 4.3E-17 Te127m 2.4E-18 Br84 1.8E-17 Te127 9.4E-18 Br85 1.2E-18 Te129m 6.8E-18 I129 1.3E-22 Te129 9.3E-18 I130 6.2E-17 Te131m 2.2E-17 I131 1.6E-15 Te131 9.8E-18 I132 7.2E-16 Te132 1.6E-16 I133 2.4E-15 Te133m 1.3E-17 I134 4.1E-16 Te134 1.8E-17 I135 1.5E-15 Ba137m 3.3E-16 Rubidium, Cesium Ba139 3.6E-19 Rb86m 4.5E-22 Ba140 2.0E-18 Rb86 1.2E-17 La140 5.8E-19 Rb88 1.7E-15 La141 1.1E-19 Rb89 7.5E-17 La142 5.3E-20 Cs132 2.3E-19 Ce141 3.1E-19 Cs134 1.7E-15 Ce143 2.3E-19 Cs135m 1.3E-18 Ce144 2.6E-19 Cs136 3.7E-16 Pr143 2.7E-19 Cs137 8.7E-16 Pr144 2.2E-19 Cs138 6.8E-16 Np239 4.9E-18 Other FPs Corrosion/Activation Products - CRUD P32 3.0E-23 Na24 4.9E-15 Co57 2.3E-25 Cr51 2.8E-16 Sr89 1.4E-18 Mn54 1.4E-16 Sr90 2.1E-19 Fe55 1.1E-16 Sr91 7.1E-19 Fe59 2.7E-17 Sr92 3.7E-19 Co58 4.1E-16 Y90 5.2E-20 Co60 4.7E-17 Y91m 3.6E-19 Ni63 2.3E-17 Y91 2.0E-19 Zn65 4.5E-17 Y92 3.2E-19 Zr95 3.5E-17 Y93 1.5E-19 Ag110m 1.2E-16 cale US460 SDAA 11.1-14 Revision 1

Secondary Coolant Secondary Coolant Nuclide Concentrations Nuclide Concentrations (Ci/g) (Ci/g)

Zr97 2.2E-19 W187 2.5E-16 Nb95 5.6E-19 Water Activation Products Mo99 4.0E-16 C14 9.4E-17 Ar41 5.0E-14 cale US460 SDAA 11.1-15 Revision 1

Primary Coolant Reactor Coolant Realistic Secondary Design Basis Average System Letdown / Coolant Secondary Coolant Recycle Mode Concentration CVCS Outlet Concentration Concentration (Ci/g) (Ci/g) (Ci/g) (Ci/g) ecycle (discharge) 1.3E-06 1.0E-06 2.4E-09 ---

ycle to reactor pool 1.3E-06 1.0E-06 2.5E-09 ---

eup ycle back to CVCS 2.6E-06 2.6E-06 --- 4.7E-09 eup

The maximum calculated peak primary coolant tritium concentration is 3.3 µCi/g.

cale US460 SDAA 11.1-16 Revision 1

The liquid waste management system is called the liquid radioactive waste system (LRWS). The LRWS is designed to collect, hold, and process liquid radioactive waste generated from normal operations and anticipated operational occurrences (AOOs). After processing and satisfactory sampling, liquids may be recycled or discharged. The LRWS is operated in a batch mode by an operator located in the waste management control room (WMCR).

The LRWS receives radioactive fluids from the chemical and volume control system (CVCS), the solid radioactive waste system (SRWS), the containment evacuation system (CES), the reactor component cooling water system (RCCWS), mixtures from the boron addition system, waste water from pool cooling and cleanup system (PCWS),

contaminated liquids from the balance-of-plant drain system, and the radioactive waste drain system (RWDS). The LRWS components are located in the Reactor Building (RXB) and in the Radioactive Waste Building (RWB).

2.1 Design Bases The LRWS has no safety-related function and is not risk significant. A failure of the LRWS does not adversely affect safety-related systems or components. Table 11.2-8 identifies SSC classifications for LRWS. The LRWS is not credited for mitigation of design basis accidents and has no safe shutdown functions. General Design Criteria (GDC) 2, 3, 60, and 61 are considered in the design of the LRWS. Section 11.2.2.6 contains further detail.

The LRWS is designed to comply with the as low as reasonably achievable (ALARA) philosophy of 10 CFR 20.1101(b) and the dose limits of 10 CFR 20.1301, 10 CFR 20.1302, and 10 CFR 50 Appendix I ALARA design objectives, including the effluent concentration limits of 10 CFR 20 Appendix B, Table 2 and 40 CFR 190 as implemented under 10 CFR 20.1301(e). A design objective of the LRWS is to provide the capability for sampling to ensure that liquid releases of radioactive material in liquid effluents are ALARA. Section 12.1 contains more detail about ways ALARA is implemented into the design.

2.2 System Description The LRWS includes tanks, pumps, filters, and ion exchangers to receive, store, process, and monitor liquid radioactive waste to be recycled or released to the environment in accordance with regulations.

The LRWS collects, processes, and releases radioactive and potentially radioactive liquid wastes produced by the plant during the plant lifecycle. Separate collection tanks are provided for low-conductivity waste subsystem (LCW), high-conductivity waste subsystem (HCW), and detergent wastes. Oily waste is removed by an oil separator, collected in drums and sent to the SRWS for eventual shipment offsite.

The remaining fluid in the separator is sent to the HCW Collection Tanks. Chemical wastes are collected as a part of the RWDS. Mixed wastes are collected locally in drums and sent offsite. The LRWS processing equipment flow path consists of two cale US460 SDAA 11.2-1 Revision 1

The liquid wastes from the various sources are temporarily stored in collection tanks located in the RWB. System equipment and components are located in stainless-steel-lined, shielded cubicles as necessary to contain leaks and for radiation shielding. Other equipment areas, located outside of steel-lined cubicles, have concrete surfaces that are sealed with a qualified coating. The system operates on a batch basis, using skid-based processing equipment that includes filters, ion exchangers, and reverse osmosis components. Subsequent to processing, the liquid is routed to sample tanks to monitor the quality of the liquid before recycling or release. If the water quality is not acceptable, the water is returned to a collection tank for further treatment.

The LRWS is designed with sufficient capacity to process liquid wastes during periods of equipment maintenance or failures and during periods of abnormal waste generation. To meet these processing demands, interconnections between LRWS components, redundant equipment, skid-based equipment, liquid holdup storage, and treatment capacity are provided in the design.

The LRWS is designed to control leakage and facilitate access, operation, inspection, testing, and maintenance to maintain radiation exposures to operating and maintenance personnel as low as is reasonably achievable and to minimize contamination of the facility.

The LRWS design includes the following maintenance considerations:

  • location of redundant permanent plant equipment in separate shielded cubicles
  • clean-in-place provisions to reduce the radiation source term before maintenance
  • redundant components allow uninterrupted waste operation and flexibility in maintenance scheduling When the sample results of the LRWS meet discharge limits, the utility water system (UWS) discharges the treated effluent to the environment. The UWS dilutes the liquid effluent further before discharge.

2.2.1 Low Conductivity Waste Subsystem The LCW consists of coolant-grade boron and hydrogen-containing wastes with high radioactivity concentrations. The CVCS letdown during normal operation and reactor heatup, along with the pressurizer vent letdown before and during reactor shutdown, are routed to the LRWS degasifiers for hydrogen and fission gas removal. The stripped gases are sent to the gaseous radioactive waste system (GRWS) for holdup and radioactive decay. The liquid waste is routed to the LCW collection tanks. The LCW collection tank fluid can be sent back to the degasifiers for additional gas removal if sample results determine it necessary. The LCW collection tanks have connections to be purged with nitrogen for inerting, as required.

cale US460 SDAA 11.2-2 Revision 1

phase separator tanks, liquid from the SRWS dewatering process, and pool water from the PCWS and out-of-specification boric acid batches from the boron addition system.

The LCW collection tanks provide for sampling of the waste on a batch basis before sending to the LCW processing equipment. The LCW processing equipment is designed to handle filtering and removal of radioactive waste. The treated liquid waste is routed to the LCW sample tanks for sampling. The treated effluent is either recycled for use within the plant or discharged to the environment through the UWS. The UWS dilutes the liquid effluent further before discharge to the environment. If the LCW sample tank sample results do not meet specified requirements, the waste can be returned to the collection tanks for reprocessing.

Reactor coolant letdown may be sent from the LCW collection tanks to the LCW sample tanks and then back to CVCS, bypassing the LCW processing equipment.

Treated liquid waste may also be routed to the HCW sample tanks or the drum dryer skid.

2.2.2 High Conductivity Waste Subsystem The HCW consists of liquid radioactive waste containing a varying degree of suspended solids and low radioactivity concentration. The HCW collection tanks collect waste from the following sources:

  • The RXB floor drains through RWDS
  • The RWB floor drains through RWDS
  • The balance-of-plant drain system chemical waste collection tank
  • The PCWS pool surge control tank overflow
  • The RWDS Reactor Building reactor component cooling water system drain tank
  • The RWDS Reactor Building chemical drain tank The HCW collection tank contents can be sampled before sending to the HCW processing equipment. The HCW processing equipment contains two carbon filter vessels. The HCW may be routed through the HCW processing equipment to the LCW processing equipment for additional treatment. Treated effluent is routed to the HCW sample tanks for sampling. When the sample result meets discharge limits, the treated effluent is discharged to the environment via UWS, or it is recycled for use within the plant. If needed, the waste can be returned to the HCW or LCW collection tanks for reprocessing.

2.2.3 Chemical Waste Processing Chemical waste is collected in the RWDS. If it is contaminated, operators send it to the HCW collection tanks. The RCCWS drains are collected separately in the RWDS to prevent the introduction of nitrite into resins that may be used in the LCW processing equipment to avoid the potential for exothermic reactions. The cale US460 SDAA 11.2-3 Revision 1

2.2.4 Detergent Waste Subsystem Detergent wastes from personnel decontamination showers and small component decontamination sinks are collected in a dedicated collection tank and sampled.

Wastes are then discharged through a cartridge filter if the sample results indicate that the specified requirements are satisfied. If the detergent sample result is not within the discharge limits, the contents go to the drum dryer skid. Detergent wastes are collected and processed separately to prevent degradation of the HCW and LCW processing equipment. There is a single train of detergent waste due to the low volume of waste generation expected, as there is no on-site laundry for the design.

2.2.5 Off-Normal Operations The LRWS is designed to be tolerant of failures and abnormal conditions as summarized in Table 11.2-2.

2.2.6 Safety Evaluation The LRWS complies with the following GDC:

  • GDC 2 as it relates to structures and components of the LRWS, by using the guidance of RG 1.143 for the seismic, safety and quality classifications.
  • GDC 3 as it relates to protecting the LRWS from the effects of fires or explosions by avoiding the generation of explosive gas mixtures and exothermic reactions with ion exchange resins.
  • GDC 60 as it relates to the design of the LRWS to control releases of radioactive liquid effluents generated during normal reactor operations, including AOOs (Section 11.2.3).
  • GDC 61 as it relates to radioactive waste systems being designed to provide for adequate safety under normal and postulated accident conditions, and designed with suitable shielding for radiation protection and with appropriate containment, confinement, and filtering systems.

The LRWS components are evaluated and classified as RW-IIa, RW-IIb or RW-IIc, as described in RG 1.143, by comparing the radioisotopic content of the component with the A1 and A2 quantities listed in Appendix A of 10 CFR 71. The safety classification for the LRWS components applies to components up to and including the nearest isolation device. The resulting safety classifications for LRWS components are listed in Table 11.2-1. The applicable standards from RG 1.143 Table 1 are used in the design, construction, and testing of the LRWS components. The applicable design criteria from RG 1.143 Table 2, Table 3, and Table 4 are used in the design analysis of the LRWS components.

cale US460 SDAA 11.2-4 Revision 1

contamination of the facility and the environment, facilitate eventual decommissioning, and reduce the generation of radioactive waste. Additional details are provided in Section 12.3.

The design of the principle components, piping, and valves that contain radioactive fluids comply with the seismic and quality requirements of RG 1.143.

The design of the LRWS utilizes and conforms to the guidance provided in RG 1.143, including Branch Technical Position 11-6.

The RWB safety classification is described in Section 3.2 and presented in Table 3.2-1.

2.3 Radioactive Effluent Releases 2.3.1 Radioactive Releases The system design reduces liquid effluent discharges from the LRWS to the environment by adequately processing liquid wastes and monitoring releases.

The design employs the use of a single point of discharge for liquid effluents to the environment through the LRWS discharge header, which is sent to the UWS discharge basin.

The calculation of liquid effluent releases is consistent with RG 1.112, as modified by Technical Report TR-123242 (Reference 11.2-1). The calculation of off-site dose consequences from normal liquid effluents is consistent with RG 1.109.

The total resultant liquid release concentrations are provided in Table 11.2-9, and demonstrate compliance with 10 CFR 20 Appendix B, Table 2.

The maximum individual doses are calculated using the LADTAP II Code, using the input parameters listed in Table 11.2-6. The resultant doses are presented in Table 11.2-7 and demonstrate compliance with the limits of 10 CFR 50 Appendix I.

L Item 11.2-1: An applicant that references the NuScale Power Plant US460 standard design will calculate doses to members of the public using the site-specific parameters, compare those liquid effluent doses to the numerical design objectives of 10 CFR 50, Appendix I, and comply with the requirements of 10 CFR 20.1302 and 40 CFR 190.

2.3.2 Compliance with Branch Technical Position 11-6 The only outdoor tank expected to contain radioactive liquids is the PCWS pool surge control storage tank, described in FSAR Section 9.1. The PCWS pool surge control storage tank secondary containment tank has sufficient volume to store the contents of the PCWS pool surge control storage tank plus the contents of related piping. The radionuclide concentration water of the PCWS pool surge cale US460 SDAA 11.2-5 Revision 1

L Item 11.2-2: An applicant that references the NuScale Power Plant US460 standard design will perform a site-specific evaluation of the consequences of an accidental release of radioactive liquid from the pool surge control storage tank in accordance with NRC Branch Technical Position 11-6.

2.3.3 Dilution Factors The utility water discharge flow rate is credited in the calculation of the discharge concentrations, as described in Reference 11.2-1. The design ensures that the discharge concentrations are within 10 CFR 20 Appendix B, Table 2, limits. The unrestricted area doses are calculated using a dilution factor as shown in Table 11.2-6, which results in the unrestricted area doses being within 10 CFR 50, Appendix I limits.

L Item 11.2-3: An applicant that references the NuScale Power Plant US460 standard design will perform a site-specific evaluation using the site-specific source term and dilution flow for liquid effluent releases, and confirm that the discharge concentrations do not exceed the limits specified by 10 CFR 20, Appendix B, Table 2.

2.3.4 Site-Specific Cost-Benefit Analysis L Item 11.2-4: An applicant that references the NuScale Power Plant US460 standard design will perform a cost-benefit analysis as required by 10 CFR 50.34a and 10 CFR 50, Appendix I, to demonstrate conformance with regulatory requirements. This cost-benefit analysis is to be performed using the guidance of Regulatory Guide 1.110.

2.4 Testing and Inspection Requirements Section 14.2 describes the LRWS preoperational tests and includes the applicable testing and inspection requirements from RG 1.143.

The design incorporates inspection and testing provisions to enable periodic evaluation of the operability and requires functional performance of active components of the system.

2.5 Instrumentation and Controls The plant controls and indications for filling waste collection tanks are automatic and are controlled by the plant control system with indication in the WMCR. The atmospheric tanks in LRWS include high-level alarms and controls to prevent overflow. If a collection tank high-level alarm is received, system valves automatically realign to direct the incoming waste flows toward the collection tank that is in the standby mode.

cale US460 SDAA 11.2-6 Revision 1

with totalizer, and leak detection that monitors the pipes annulus. The double-walled pipes annulus is pressurized to be greater than the process or groundwater pressure and is alarmed to stop the discharge flow upon an indication of low pressure.

A liquid radioactive waste discharge automatically isolates upon an alarm due to a low dilution flow indication, a low pressure indication in the discharge pipe annulus, or a high-radiation alarm in a discharge line radiation monitor.

2.6 Reference 11.2-1 NuScale Power, LLC, Effluent Release (GALE Replacement)

Methodology and Results, TR-123242, Revision 1.

cale US460 SDAA 11.2-7 Revision 1

Scale Final Safety Analysis Report Component RG 1.143 Type Capacity Design Design Material Table for Assumed (Quantity) Safety Pressure Temperature Radioactive Content Classification (psig) (°F) asifier (2) RW-IIa Vertical 12,500 gallons 150 550 Stainless Steel Table 12.2-15a asifier Liquid Transfer Pumps (2) RW-IIc Sealless Centrifugal 28 gpm 150 210 Stainless Steel -

collection tank (2) RW-IIc Vertical, conical 16,000 gallons 15 240 Stainless Steel Table 12.2-12a collection tank transfer pump (2) RW-IIc Sealless Centrifugal 39 gpm 290 155 Stainless Steel -

collection tank (2) RW-IIc Vertical Conical 16,000 gallons 15 200 Stainless Steel Table 12.2-12a collection tank transfer pump (2) RW-IIc Sealless Centrifugal 39 gpm 230 155 Stainless Steel -

sample tank (2) RW-IIc Vertical conical 16,000 gallons 15 155 Stainless Steel Table 12.2-12b sample tank transfer pump (2) RW-IIc Sealless Centrifugal 28 gpm 150 155 Stainless Steel -

sample tank (2) RW-IIc Vertical conical 16,000 gallons 15 155 Stainless Steel Table 12.2-12b sample tank transfer pump (2) RW-IIc Sealless Centrifugal 28 gpm 150 155 Stainless Steel -

eparator (1) RW-IIc - 240 gpm 150 155 Stainless Steel Table 12.2-12a rgent waste collection tank (1) RW-IIc Vertical conical 500 gallons 15 200 Stainless Steel -

rgent waste drain filter (1) RW-IIc Cartridge 20 micron 150 155 Stainless Steel -

ineralized water break tank (1) RW-IIc Vertical 10,000 gallons 15 155 Stainless Steel -

Processing charcoal filter (2) RW-IIb Vertical Vessel 35 gpm 230 155 Stainless Steel LCW- Table 12.2-12a HCW- Table 12.2-12b Reverse Osmosis Skid (1) RW-IIc Vertical 35 gpm 290 155 Stainless Steel LCW- Table 12.2-12a HCW- Table 12.2-12b n-In-Place Skid (1) RW-IIc - 55 gallons 150 155 Stainless Steel -

m Dryer Skids (1) RW-IIc - 55 gpd 15 155 Stainless Steel -

Pre-conditioning filter vessels (2) RW-IIa - 35 gpm 290 155 Stainless Steel Liquid Waste Management System accumulator vessels (3) RW-IIa - 35 gpm 290 155 Stainless Steel Ion Exchange vessel (5) RW-IIa - 35 gpm 290 155 Stainless Steel Table 12.2-12a solids collection filter (1) RW-IIc - 35 gpm 290 155 Stainless Steel Polishing Ion Exchange Vessel (4) RW-IIc - 35 gpm 290 155 Stainless Steel Table 12.2-12a ineralized water break tank transfer RW-IIc Sealless centrifugal 220 gpm 150 155 Stainless Steel -

p (1)

Consequences Off Normal Operation/AOO Consequences Event Indication System Response Corrective Action level or loss of Automatic Switch to standby degasifier Corrective maintenance of idle um in Degasifier equipment n in use asifier transfer Automatic Switch to standby degasifier Corrective maintenance of idle p trips equipment ection Tank Automatic Switch to standby Collection Tank Corrective maintenance of idle sfer Pump failure Transfer Pump equipment

-high level in Operator surveillance Discharge valves automatically Investigate cause and procedures ection or Sample close and runnning pump stops to prevent challenges to control ks system train of Operator surveillance Based on sample results, use one Repair or replace components essing equipment train to alternately process LCW and restore operability erable and HCW ple Tank shows Review of sample Pump to Collection Tank for Diagnose cause and repair or ple of high results reprocessing replace as necessary oactivity radiation on Automatic Discharge valves automatically Diagnose cause, reprocess le point LRW close and runnning pump stops remaining sample tank contents harge guard pipe Automatic Discharge valves automatically Diagnose cause, repair buried sure on buried close and running LRW discharge pipe discharge pipe radiation alarm Local and MCR alarm pump stops Investigate cause and initiate cleanup and corrective maintenance ks or spills Operator surveillance Suspend processing, prevent the Investigate cause and initiate (or leak detection spread of contamination cleanup and corrective alarm or area radiation maintenance alarm) of Nitrogen Indication from vendor Vendor control system isolates Investigate, restore Nitrogen sure to control system degasifier skid pressure and resume operation asifiers cale US460 SDAA 11.2-9 Revision 1

LRWS Input Source Expected Input Expected Activity Rate (6 NPMs)

(gpy) collection tank RXB/RWB equipment drains 2.9E+04 0.001 primary coolant activity (PCA)

Other equipment drains 1.1E+04 0.093 PCA Normal letdown (six operating units) 1.9E+05 CVCS outlet Degasification prior to shutdown (six events per year) 3.0E+03 primary coolant through evaporator Additional CVCS letdown streams 3.8E+04 CVCS outlet LCW Total 2.7E+05 W collection tank RXB/RWB floor drains (via oil separator) 7.3E+04 0.1 PCA RXB RCCW drain tank (via oil separator) 3.6E+01 0.001 PCA Hot machine shop, decontamination room sump (via oil 9.0E+04 0.01 PCA eparator)

RXB chemical drain tank (Hot lab sink) (via oil separator) 4.4E+03 0.05 PCA RXB chemical drain tank (CES sample tank) (via oil 2.2E+04 primary coolant through eparator) evaporator Pump seal leaks (via oil separator) 8.1E+03 0.1 PCA Valve packing leaks (via oil separator) 4.8E+03 0.1 PCA Groundwater / Condensation (via oil separator) 2.5E+05 0.001 PCA Equipment area decontamination (outside hot machine shop) 1.5E+04 0.01 PCA via oil separator)

Secondary coolant sampling drains 4.2E+03 Secondary coolant Condensate polisher rinse and transfer 3.6E+04 Secondary coolant Condensate polisher regeneration solutions 1.0E+04 Secondary coolant Turbine Generator Building floor drains 2.2E+04 Secondary coolant Pool water source streams 2.9E+05 Pool water source term CVCS outlet sources 3.5E+04 CVCS outlet HCW Total 8.6E+05

Assumes six NPMs operating on an 18-month refueling cycle.

cale US460 SDAA 11.2-10 Revision 1

NuScale Effluent Source Term Model Assumption Value Units ary coolant source term Table 11.1-6 S demineralizer decontamination factors-logens 100 --

, Rb 2 --

hers 50 --

l water source Table 12.2-9 liquid partition fractions-and PCA through Evaporator ble gases 1 --

logens 100 --

hers 1000 --

ondary coolant source term Table 11.1-7 adjustment 0.07 Ci/year S dilution factor for 10 CFR 20 Appendix B 700 gpm cale US460 SDAA 11.2-11 Revision 1

Header LRWS LCW Sample LRWS HCW Sample Plant Liquid Release Total Liquid Release Tank Release Tank Release without AOO Adjustment with AOO Adjustment uclide (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) 2 1.4E-12 2.3E-09 2.3E-09 4.9E-08 3 - 2.5E-30 2.5E-30 5.3E-29 4.9E-13 1.4E-12 1.9E-12 4.2E-11 1.2E-19 2.1E-11 2.1E-11 4.6E-10 1.0E-06 1.6E-05 1.7E-05 3.7E-04 1.6E-08 3.6E-07 3.8E-07 8.2E-06 4.5E-13 5.8E-08 5.8E-08 1.3E-06 4.1E-29 5.3E-14 5.3E-14 1.1E-12 6 3.6E-06 1.4E-06 5.0E-06 1.1E-04 32 1.6E-08 1.6E-08 3.3E-08 7.1E-07 34 1.1E-03 2.7E-04 1.4E-03 3.0E-02 36 8.1E-05 3.9E-05 1.2E-04 2.6E-03 37 5.8E-04 1.4E-04 7.2E-04 1.6E-02 9.8E-14 2.4E-13 3.4E-13 7.3E-12 7 2.0E-15 2.6E-15 4.5E-15 9.9E-14 9 1.9E-08 1.7E-08 3.6E-08 7.8E-07 0 1.9E-09 2.5E-09 4.4E-09 9.5E-08 1 1.0E-24 1.3E-14 1.3E-14 2.8E-13 2 - 4.1E-30 4.1E-30 9.0E-29 1.9E-09 2.2E-09 4.1E-09 9.0E-08 m 6.5E-25 8.1E-15 8.1E-15 1.8E-13 1.4E-09 2.1E-09 3.5E-09 7.7E-08

- 2.7E-24 2.7E-24 5.8E-23 1.6E-24 5.6E-15 5.6E-15 1.2E-13 7 1.7E-18 1.2E-12 1.2E-12 2.6E-11 5 9.7E-08 6.7E-07 7.7E-07 1.7E-05 9 1.8E-08 6.6E-07 6.8E-07 1.5E-05 9m 1.7E-08 6.4E-07 6.6E-07 1.4E-05 9 6.9E-11 8.9E-11 1.6E-10 3.4E-09 03 2.4E-09 3.9E-09 6.3E-09 1.4E-07 05 - 3.7E-22 3.7E-22 8.1E-21 06 2.1E-09 2.8E-09 4.9E-09 1.1E-07 03m 2.4E-09 3.9E-09 6.3E-09 1.4E-07 05 1.3E-13 8.5E-11 8.5E-11 1.8E-09 06 2.1E-09 2.8E-09 4.9E-09 1.1E-07 10 1.3E-08 1.8E-07 1.9E-07 4.2E-06 24 4.1E-12 6.1E-12 1.0E-11 2.2E-10 25 3.9E-11 5.0E-11 8.8E-11 1.9E-09 27 4.4E-12 6.2E-11 6.7E-11 1.5E-09 29 - 6.0E-23 6.0E-23 1.3E-21 25m 4.4E-09 6.6E-09 1.1E-08 2.4E-07 27m 1.9E-08 2.6E-08 4.5E-08 9.8E-07 27 1.8E-08 2.6E-08 4.4E-08 9.6E-07 29m 4.0E-08 6.7E-08 1.1E-07 2.3E-06 cale US460 SDAA 11.2-12 Revision 1

LRWS LCW Sample LRWS HCW Sample Plant Liquid Release Total Liquid Release Tank Release Tank Release without AOO Adjustment with AOO Adjustment uclide (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) 29 2.5E-08 4.2E-08 6.8E-08 1.5E-06 31m 1.7E-12 3.5E-09 3.5E-09 7.6E-08 31 3.9E-13 7.9E-10 7.9E-10 1.7E-08 32 1.5E-08 3.5E-07 3.7E-07 8.0E-06 37m 5.5E-04 1.3E-04 6.8E-04 1.5E-02 40 5.8E-09 1.5E-08 2.1E-08 4.5E-07 40 6.6E-09 1.7E-08 2.3E-08 5.1E-07 41 - 7.1E-24 7.1E-24 1.5E-22 41 1.8E-09 3.0E-09 4.8E-09 1.0E-07 43 5.2E-14 5.4E-11 5.4E-11 1.2E-09 44 2.2E-09 2.9E-09 5.2E-09 1.1E-07 43 9.2E-10 2.3E-09 3.2E-09 7.0E-08 44 2.2E-09 2.9E-09 5.1E-09 1.1E-07 39 9.0E-11 5.8E-09 5.8E-09 1.3E-07 4 3.0E-15 1.0E-08 1.0E-08 2.2E-07 1 1.5E-06 2.5E-05 2.7E-05 5.8E-04 4 1.2E-06 1.6E-05 1.7E-05 3.8E-04 5 9.5E-07 1.2E-05 1.3E-05 2.9E-04 9 1.7E-07 2.7E-06 2.9E-06 6.2E-05 8 3.0E-06 3.9E-04 4.0E-04 8.7E-03 0 4.2E-07 5.5E-06 5.9E-06 1.3E-04 3 2.1E-07 2.7E-06 3.0E-06 6.4E-05 5 3.9E-07 5.1E-06 5.5E-06 1.2E-04 5 2.5E-07 3.6E-06 3.9E-06 8.5E-05 10m 9.9E-07 1.3E-05 1.4E-05 3.1E-04 87 8.7E-13 3.6E-08 3.6E-08 7.8E-07 8.6E+02 3.0E+02 1.2E+03 1.2E+03 al 8.6E+02 3.0E+02 1.2E+03 1.2E+03 cale US460 SDAA 11.2-13 Revision 1

Parameter Value Units rce term Table 11.2-5 --

re-width factor 1.0 --

charge flow rate 605 gpm oundment reconcentration model None --

ation rate 100 liters/m2-month tion factor for aquatic food, boating, shoreline, swimming and drinking water 1 --

tion factor for irrigation water usage location 1 --

type Freshwater --

osure Pathway- --

ansit time - aquatic food 0 --

ansit time - boating 0 --

ansit time - swimming 0 --

ansit time - shoreline 0 --

ansit time - drinking water 0 --

ansit time - irrigated crops 0 --

ansit time - milk/meat animal water usage 0 --

ction of crops irrigated using non-contaminated water 0 --

ction of milk/meat animal feed irrigated using non-contaminated water 0 --

ction of milk/meat animal drinking water from non-contaminated water 0 --

imum dilution flow rate 141 cfs cale US460 SDAA 11.2-14 Revision 1

10 CFR 50, Appendix I ALARA Calculated Dose Design Objective Type of Dose (mrem/yr) (mrem/yr) l Body 2.9 3 idual Organ 4.5 10 cale US460 SDAA 11.2-15 Revision 1

Scale Final Safety Analysis Report SSC (Note 1) Location SSC Augmented Design Quality Group/Safety Seismic Classification Classification Requirements Classification (Ref RG (Ref. RG 1.29 or RG (A1, A2, B1, B2) (Note 2) 1.26 or RG 1.143) 1.143) (Note 4)

(Note 3)

LRWS, Liquid Radioactive Waste System omponents (except those listed below): RWB/RXB B2 RG 1.143 RW-IIc III gasifier (including condensers & vacuum RWB/RXB B2 RG 1.143 RW-IIa RW-IIa mps)

W accumulator, ion exchanger &

e-condition filter vessels W processing equipment RWB B2 RG 1.143 RW-IIb III 1: Acronyms used in this table are listed in Table 1.1-1 2: Additional augmented design requirements, such as the application of a Quality Group, Radwaste safety, or seismic classification, to nonsafety-related SSC are reflected in the columns Quality Group / Safety Classification and Seismic Classification, where applicable. Environmental Qualifications of SSC are identified in Table 3.11-1.

3: Section 3.2.2.1 through Section 3.2.2.4 provides the applicable codes and standards for each RG 1.26 Quality Group designation (A, B, C, and D). A Quality Group classification per RG 1.26 is not applicable to supports or instrumentation. Section 3.2.1.4 provides a description of RG 1.143 classification for RW-IIa, RW-IIb, and RW-IIc.

4: Where SSC (or portions thereof) as determined in the as-built plant that are identified as Seismic Category III in this table could, as the result of a seismic event, adversely affect Seismic Category I SSC or result in incapacitating injury to occupants of the control room, they are categorized as Seismic Category II consistent with Section 3.2.1.2 and analyzed as described in Section 3.7.3.8.

Liquid Waste Management System

Limits Discharge Concentration Concentration Limit Nuclide (µCi/ml) (µCi/ml) Fraction of Limit 3.5E-14 4.0E-05 8.8E-10 3.0E-17 2.0E-07 1.5E-10 3.3E-16 2.0E-05 1.6E-11 2.7E-10 1.0E-06 2.7E-04 5.9E-12 1.0E-04 5.9E-08 9.1E-13 7.0E-06 1.3E-07 8.2E-19 3.0E-05 2.7E-14 6 7.8E-11 7.0E-06 1.1E-05 32 5.1E-13 4.0E-05 1.3E-08 34 2.1E-08 9.0E-07 2.4E-02 36 1.9E-09 6.0E-06 3.1E-04 37 1.1E-08 1.0E-06 1.1E-02 7 7.1E-20 6.0E-05 1.2E-15 5.6E-13 8.0E-06 7.0E-08 6.8E-14 5.0E-07 1.4E-07 2.0E-19 2.0E-05 9.9E-15 6.4E-14 7.0E-06 9.2E-09 m 1.3E-19 2.0E-03 6.3E-17 5.5E-14 8.0E-06 6.9E-09 4.2E-29 4.0E-05 1.0E-24 8.7E-20 2.0E-05 4.4E-15 1.8E-17 9.0E-06 2.1E-12 5 1.2E-11 3.0E-05 4.0E-07 9 1.1E-11 2.0E-05 5.3E-07 9m 1.0E-11 1.0E-03 1.0E-08 9 2.5E-15 6.0E-05 4.1E-11 03 9.9E-14 3.0E-05 3.3E-09 05 5.8E-27 7.0E-05 8.3E-23 06 7.6E-14 3.0E-06 2.5E-08 03m 9.8E-14 6.0E-03 1.6E-11 05 1.3E-15 5.0E-05 2.6E-11 24 1.6E-16 7.0E-06 2.3E-11 25 1.4E-15 3.0E-05 4.6E-11 27 1.0E-15 1.0E-05 1.0E-10 29 9.4E-28 4.0E-05 2.3E-23 25m 1.7E-13 2.0E-05 8.6E-09 27m 7.0E-13 9.0E-06 7.8E-08 27 6.9E-13 1.0E-04 6.9E-09 29m 1.7E-12 7.0E-06 2.4E-07 29 1.1E-12 4.0E-04 2.6E-09 31m 5.5E-14 8.0E-06 6.8E-09 31 1.2E-14 8.0E-05 1.5E-10 32 5.7E-12 9.0E-06 6.4E-07 40 3.3E-13 8.0E-06 4.1E-08 40 3.6E-13 9.0E-06 4.0E-08 cale US460 SDAA 11.2-17 Revision 1

Discharge Concentration Concentration Limit Nuclide (µCi/ml) (µCi/ml) Fraction of Limit 41 1.1E-28 5.0E-05 2.2E-24 41 7.5E-14 3.0E-05 2.5E-09 43 8.5E-16 2.0E-05 4.2E-11 44 8.1E-14 3.0E-06 2.7E-08 3 5.0E-14 2.0E-05 2.5E-09 4 8.0E-14 6.0E-04 1.3E-10 39 9.1E-14 2.0E-05 4.6E-09 4 1.6E-13 5.0E-05 3.2E-09 1 4.2E-10 5.0E-04 8.4E-07 4 2.7E-10 3.0E-05 9.1E-06 5 2.1E-10 1.0E-04 2.1E-06 9 4.5E-11 1.0E-05 4.5E-06 8 6.2E-09 2.0E-05 3.1E-04 0 9.2E-11 3.0E-06 3.1E-05 4.6E-11 1.0E-04 4.6E-07 5 8.6E-11 5.0E-06 1.7E-05 6.1E-11 2.0E-05 3.0E-06 10m 2.2E-10 6.0E-06 3.7E-05 7 5.6E-13 3.0E-05 1.9E-08 8.3E-04 1.0E-03 8.3E-01 l 8.3E-04 1.3E-02 8.6E-01 cale US460 SDAA 11.2-18 Revision 1

NuScale Final Safety Analysis Report Liquid Waste Management System Figure 11.2-1a: Liquid Radioactive Waste System Diagram TO RBVS PIT FROM CHILLED FROM DEMINERALIZED FROM CHILLED WATER SUPPLY WATER WATER SUPPLY TIT FROM LCW (B) (B) TE LRWS DEGASIFIER SKID A LRWS DEGASIFIER SKID B COLLECTION TANKS (A) BBY1 TO GRWS (C)

FROM CVCS LETDOWN (A) (C) TO CHILLED WATER RETURN FROM CVCS PRESSURIZER VENT TO CHILLED WATER RETURN FROM NITROGEN DISTRIBUTION SYSTEM LS LS PIT TO LCW COLLECTION TANKS PIT LRWS DEGASIFIER B LIQUID TRANSFER PUMP LRWS DEGASIFIER A LIQUID TRANSFER PUMP NuScale US460 SDAA 11.2-19 Revision 1

NuScale Final Safety Analysis Report Liquid Waste Management System Figure 11.2-1b: Liquid Radioactive Waste System Diagram FROM NITROGEN TO RWBVS DISTRIBUTION SYSTEM FROM SRWS (DEWATERING SKID INFLUENT)

FROM PCWS (INFLUENT)

LCW COLLECTION TANK A FROM DEGASIFIER LIT FROM BORON ADDITION SYSTEM (INFLUENT) PIT LS FROM LCW SAMPLE TANKS LRWS LCW COLLECTION TANK A TO LCW TO PROCESSING SKID TRANSFER PUMP RWBVS FROM RADIOACTIVE WASTE DRAIN SYSTEM (RXB EQUIPMENT BBD1 DRAIN SUMPS)

LCW TO DEGASIFIERS FROM RADIOACTIVE WASTE COLLECTION DRAIN SYSTEM (RWB EQUIPMENT TANK B DRAIN SUMPS)

FROM SRWS (SPENT RESIN STORAGE LIT TANK INFLUENT)

TO LCW SAMPLE TANKS PIT LS FROM SRWS (PHASE SEPARATOR TANK INFLUENT)

LRWS LCW COLLECTION TANK B TRANSFER PUMP NuScale US460 SDAA 11.2-20 Revision 1

NuScale Final Safety Analysis Report Liquid Waste Management System Figure 11.2-1c: Liquid Radioactive Waste System Diagram PDT TO RWBVS TO RWBVS LIT FROM RWDS (RXB FLOOR DRAIN SUMP INFLUENT) HCW COLLECTION TANK A FROM RWDS OIL SEPARATOR (RWB FLOOR DRAIN SUMP INFLUENT)

LIT PIT TO HCW TO SRWS PROCESSING (55-GALLON WASTE DRUM) SKID LS LRWS HCW FROM LCW SAMPLE TANKS COLLECTION TANK A TRANSFER PUMP TO FROM BPDS RWBVS (CHEMICAL WASTE COLLECTION TANK INFLUENT)

PDT FROM PCWS (POOL SURGE INFLUENT)

HCW COLLECTION TANK B LIT PIT LS LRWS HCW COLLECTION TANK B TRANSFER PUMP NuScale US460 SDAA 11.2-21 Revision 1

NuScale Final Safety Analysis Report Liquid Waste Management System Figure 11.2-1d: Liquid Radioactive Waste System Diagram FROM DEMINERALIZED WATER SYSTEM DW BREAK TANK LIT LRWS PIT CLEAN-IN-PLACE CIP SKID SUPPLY FIT DW BREAK TANK TRANSFER PUMP A FE TO SRWS (SPENT RESIN STORAGE AND PIT PHASE SEPARATOR TANKS)

TO PCWS (DEMINERALIZERS)

DW BREAK TANK TRANSFER PUMP B TO DRUM DRYER SKID NuScale US460 SDAA 11.2-22 Revision 1

NuScale Final Safety Analysis Report Liquid Waste Management System Figure 11.2-1e: Liquid Radioactive Waste System Diagram FROM SERVICE AIR TO HCW PROCESSING SKID TO DRUM LRWS LCW PROCESSING SKID DRYER SKID FROM LCW COLLECTION TANKS TO RWBVS PIT FROM HCW TO LCW PROCESSING SKID SAMPLE TANKS FROM SRWS TO HCW (PHASE SEPARATOR TANK SAMPLE TANKS TRANSFER PUMP)

TO SRWS (PHASE SEPARATOR TANKS)

NuScale US460 SDAA 11.2-23 Revision 1

NuScale Final Safety Analysis Report Liquid Waste Management System Figure 11.2-1f: Liquid Radioactive Waste System Diagram TO DRUM DRYER SKID LRWS HCW PROCESSING SKID FROM SERVICE AIR TO RWBVS FROM HCW COLLECTION TANKS PIT TO HCW SAMPLE TANKS FROM SRWS (PHASE SEPARATOR TANK TRANSFER PUMPS)

TO LCW PROCESSING SKID TO SRWS (PHASE SEPARATOR TANKS)

NuScale US460 SDAA 11.2-24 Revision 1

NuScale Final Safety Analysis Report Liquid Waste Management System Figure 11.2-1g: Liquid Radioactive Waste System Diagram TO RWBVS LCW SAMPLE TANK A LIT FROM LCW PROCESSING SKID TRANSFER PUMP PIT LS FROM LCW COLLECTION TANKS LRWS LCW SAMPLE TO PCWS TANK TRANSFER PUMP A (ULTIMATE HEAT SINK)

TO RWBVS TO CVCS (MAKEUP PUMP)

TO LCW LCW COLLECTION TANKS SAMPLE TANK B TO HCW COLLECTION TANKS LIT PIT LS FQT RIT RIT PIT RE RE FE TO UTILITY WATER SYSTEM (EFFLUENT DISCHARGE)

LRWS LCW SAMPLE TANK TRANSFER PUMP B FROM HCW SAMPLE TANK TRANSFER PUMP FROM DETERGENT WASTE COLLECTION TANK TRANSFER PUMP NuScale US460 SDAA 11.2-25 Revision 1

NuScale Final Safety Analysis Report Liquid Waste Management System Figure 11.2-1h: Liquid Radioactive Waste System Diagram TO RWBVS HCW SAMPLE TANK A FROM HCW PROCESSING SKID TRANSFER PUMP LIT FROM LCW PROCESSING SKID TRANSFER PUMP PIT LS TO RWBVS LRWS HCW SAMPLE TANK TRANSFER PUMP A HCW SAMPLE TANK B TO LCW SAMPLE TANK OUTLET LIT PIT LS LRWS HCW SAMPLE TANK TRANSFER PUMP B NuScale US460 SDAA 11.2-26 Revision 1

NuScale Final Safety Analysis Report Liquid Waste Management System Figure 11.2-1i: Liquid Radioactive Waste System Diagram TO RWBVS FROM RADIOACTIVE WASTE DRAINS (RWB DECONTAMINATION SHOWER AND SINK)

DETERGENT WASTE COLLECTION TANK LIT PDT PIT TO LCW SAMPLE TANK OUTLET LRWS DETERGENT WASTE COLLECTION TANK LRWS DETERGENT TRANSFER PUMP WASTE DRAIN FILTER TO DRUM DRYER SKID NuScale US460 SDAA 11.2-27 Revision 1

NuScale Final Safety Analysis Report Liquid Waste Management System Figure 11.2-1j: Liquid Radioactive Waste System Diagram TO RWBVS FROM DW BREAK TANK FROM DETERGENT WASTE COLLECTION TANK DRUM DRYER SKID FROM LCW PROCESSING SKID FROM HCW PROCESSING SKID NuScale US460 SDAA 11.2-28 Revision 1

The gaseous radioactive waste system (GRWS) design processes the gaseous waste stream from the liquid radioactive waste system (LRWS) degasifier and the containment evacuation system (CES), provide holdup for radioactive decay of xenon and krypton, and convey the gaseous effluent to the Radioactive Waste Building heating ventilation and air conditioning (HVAC) system (RWBVS), which transports the effluent to the Reactor Building HVAC system (RBVS) for monitoring and release. The GRWS filters out particulate carryover and delays the noble gases through activated charcoal beds until they have decayed sufficiently to allow release to the environment. Design and performance of the charcoal delay system is in accordance with NUREG-0017, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)," April 1985, as modified by TR-123242 (Reference 11.3-1).

Exhaust flow from the RWBVS and RBVS are combined and monitored by the RBVS exhaust stack radiation effluent monitor before release to the environment (Section 11.5).

Primary gaseous effluent sources, besides gaseous radioactive waste, include the CES (Section 9.3.6), RWBVS (Section 9.4.3) and other sources exhausted by the RBVS (Section 9.4.2). In addition, small releases that occur in the Turbine Generator Building from the condenser air removal system (CARS) (Section 10.4.2) and turbine gland sealing system (Section 10.4.3) are monitored, but directly released to the environment.

3.1 Design Bases The GRWS serves no safety function and is not risk-significant. Table 11.3-10 identifies SSC classifications for GWMS. The GRWS does not mitigate design basis accidents and has no safe shutdown functions. General Design Criteria (GDC) 2, 3, 60, and 61 were considered in the design of the GRWS.

3.2 System Description The GRWS is in the RWB and is a passive, once-through, ambient temperature charcoal delay system that receives hydrogen-bearing gas containing fission gases from the LRWS degasifier. The GRWS also receives gaseous waste inputs from the individual NuScale Power Modules (NPMs) via the CES, if high radiation is detected in the CES exhaust. The GRWS filters particulate carryover, removes moisture, delays the gas to allow radioactive decay, and conveys it to the RBVS via the RWBVS for release to the environment through the plant exhaust stack as a monitored release (Section 11.5).

Nitrogen from the nitrogen distribution system dilutes the waste gas input from the liquid radioactive waste degasifier (and potentially CES) to maintain a hydrogen concentration of less than 4 percent. Because the waste gas input flow is not constant, the nitrogen supply maintains a positive GRWS pressure and a constant flow. The waste gas input into the GRWS passes through a vapor condenser package assembly that contains a waste gas cooler (cooled by chilled water) and a moisture separator. The moisture separator includes level control drain valves piped to the equipment drain sump in the radioactive waste drain system (RWDS). The drain line passes through a drain trap to prevent radioactive gas from passing to the RWDS in cale US460 SDAA 11.3-1 Revision 1

manual sample port. If high oxygen levels are detected, the inlet stream to the GRWS automatically isolates and a nitrogen purge flushes the GRWS. Operators manually initiate termination of nitrogen flushing and restart of normal operations.

The waste gas passes through a charcoal guard bed located in an ambient temperature-controlled shielded cubicle. Because the guard bed is at ambient room temperature, the guard bed warms the gas from the gas cooler (lowering its relative humidity) to improve fission gas capture efficiency in the decay beds. The guard bed also acts as a backup moisture-removal device. The guard bed contains a safety relief valve, differential pressure instrumentation, and a means to dry or replace charcoal. Operators manually initiate charcoal drying using remotely-operated valves and a normally deenergized charcoal drying heater, which provide a heated nitrogen flow to the guard bed. The heated, moisture-laden nitrogen recycles back to the inlet of the vapor condenser. The guard bed also contains a fire detector that automatically activates a nitrogen purge upon detecting a fire.

The conditioned waste gas then flows into either one of two charcoal decay beds, each decay bed consisting of four charcoal vessels connected in series. Entrance into the first vessel and exit from the last vessel is through the top of the vessel to minimize the potential of charcoal loss. Each decay bed contains activated charcoal optimized for xenon and krypton retention. Like the guard bed, the decay beds contain differential pressure instrumentation, fire detection instrumentation, safety relief valves, and the ability to either dry or replace charcoal. In addition, the decay beds contain radiation monitors that automatically isolate flow in the event of a high radiation indication.

The processed waste gas goes to the RWBVS, which interfaces with the RBVS that monitors the effluent path to the environment. The GRWS outlet also has an offline radiation monitor with the capability to take samples before being sent to the ventilation systems.

The gaseous radioactive waste process design is illustrated in Figure 11.3-1.

Table 11.3-1 provides the GRWS design parameters.

3.2.1 Component Description This section describes the key GRWS equipment. Table 11.3-2 summarizes specific component design parameters. Design codes, standards, and materials for construction of these components are consistent with RG 1.143, Table 1.

3.2.1.1 Waste Gas Cooler The waste gas cooler is a stainless steel, double-pipe heat exchanger that cools the incoming waste gas stream from the LRWS and CES. Chilled water (shell side) cools the waste gas stream (tube side) to condense water vapor from the gas stream to protect the charcoal beds from moisture.

cale US460 SDAA 11.3-2 Revision 1

The stainless steel moisture separator collects condensed water from the waste gas cooler. Level instrumentation controls the outlet drain valve. The condensate goes to the equipment drain waste sump in the RWDS. The drain line passes through a drain trap to prevent radioactive gases from passing to the RWDS.

3.2.1.3 Charcoal Guard Bed The charcoal guard bed is an American Society of Mechanical Engineers (ASME)Section VIII stainless steel vessel located in an ambient temperature-controlled cubicle that warms the waste gas stream, thus reducing its relative humidity. The guard bed also removes additional moisture in the waste gas stream to improve fission gas capture efficiency and protect the charcoal decay beds. The charcoal guard bed includes a safety relief valve, differential pressure instrumentation, a fire detector, and a means to dry with a charcoal drying heater or replace the charcoal, if needed.

3.2.1.4 Charcoal Decay Beds The two charcoal decay beds each consist of four ASME Section VIII stainless steel decay vessels connected in series. The vessels contain activated charcoal to allow the waste gas radionuclides to decay sufficiently before being released. Each decay bed train has a pressure relief valve, differential pressure instrumentation, and a fire detector that upon sensing a fire automatically activates a nitrogen purge. The exit of each of the two decay beds has a radiation monitor that automatically isolates the bed in the event of a high radiation signal.

3.2.1.5 Charcoal Drying Heater If needed, the charcoal drying heater is a manually initiated, stainless steel electric heater that heats nitrogen gas from the nitrogen distribution system to flow through the charcoal guard bed to dry the charcoal. Operators can also send heated nitrogen to the charcoal decay beds. After exiting the guard bed or decay beds, the nitrogen is routed back to the inlet of the waste gas cooler to remove the moisture. The charcoal drying heater has a temperature controller with a high temperature cutoff. If a fire is detected in a charcoal bed, the heater is automatically deenergized.

3.2.1.6 Oxygen and Hydrogen Analyzers There are a total of three independent oxygen analyzers and two hydrogen analyzers that continuously monitor the GRWS. Two redundant oxygen analyzers and two redundant hydrogen analyzers are downstream of the moisture separator, upstream of the charcoal guard bed, and indicate and annunciate locally, in the main control room (MCR), and in the WMCR. In the event that high oxygen levels exceed 1 percent, the system initiates an alarm locally and in both the WMCR and MCR. If the oxygen level reaches cale US460 SDAA 11.3-3 Revision 1

hydrogen monitor ensures detection of a maximum concentration of 4 percent with notification of a high-high alarm. The high alarm at approximately one-half of the maximum oxygen concentration includes a local, WMCR and MCR notification.

The design of the gas analyzer instruments is to be non-sparking. Gas analyzers have sensor checks, functional checks, and calibrations performed in accordance with vendor recommendations.

3.2.2 Malfunction Analysis Table 11.3-3 provides a summary of a malfunction analysis of the GRWS.

3.2.3 Design Safety Evaluation The GRWS complies with the following GDC found in 10 CFR Part 50, Appendix A:

  • GDC 2 as it relates to structures and components of the GRWS using the guidance of RG 1.143 for the seismic, safety, and quality classifications
  • GDC 3 as it relates to protecting the GRWS from the effects of a detonation of a hydrogen-oxygen mixture by preventing such mixtures from occurring
  • GDC 60 as it relates to the design of the GRWS to control releases of radioactive gaseous effluents generated during normal reactor operations, including AOOs
  • GDC 61 as it relates to radioactive waste systems being designed to provide for adequate safety under normal and postulated accident conditions, and designed with suitable shielding for radiation protection and with appropriate containment, confinement, and filtering systems There are design features that comply with the requirements of 10 CFR 20.1406 following the guidance of RG 4.21, to minimize contamination of the facility and the environment, facilitate eventual decommissioning, and minimize the generation of radioactive waste. Section 12.3.6 provides additional details.

The gaseous radioactive waste structures, systems, and components design complies with the codes and standards provided in RG 1.143, Table 1 through 4.

The applicable design criteria from RG1.143, Table 2, Table 3 and Table 4 are used in the design analysis of the GRWS components. The safety classification for the GRWS components applies to components, up to and including the nearest isolation device. Table 11.3-2 provides the design parameters of major components, including safety classification and operating conditions.

cale US460 SDAA 11.3-4 Revision 1

L Item 11.3-1: An applicant that references the NuScale Power Plant US460 standard design will perform a site-specific cost-benefit analysis using the guidance in Regulatory Guide 1.110.

3.2.5 Seismic Design The gaseous radioactive waste equipment and piping classification complies with RG 1.143. Section 3.7 describes the RWB seismic design.

3.3 Radioactive Effluent Releases Technical Report TR-123242 (Reference 11.3-1) describes the gaseous radioactive effluent release methodology, inputs, and results.

Table 11.3-5 tabulates the results of the radioactive effluent calculation and demonstrate compliance with the limits from 10 CFR 20, Appendix B, Table 2.

Table 11.3-4 provides the inputs. The comparison demonstrates that the overall expected gaseous releases are within the release limits.

The GASPAR II Code is used to calculate the maximum individual doses at the exclusion area boundary. Table 11.3-6 tabulates the input parameters. Table 11.3-7 tabulates the resultant doses and demonstrates compliance with the limits of 10 CFR 50 Appendix I.

L Item 11.3-2: An applicant that references the NuScale Power Plant US460 standard design will calculate doses to members of the public using the site-specific parameters, compare those gaseous effluent doses to the numerical design objectives of 10 CFR 50, Appendix I, and comply with the requirements of 10 CFR 20.1302 and 40 CFR 190.

3.3.1 Radioactive Effluent Releases and Dose Calculation due to Gaseous Radioactive Waste System Leak or Failure The analysis of a GRWS leak or failure follows the guidance of Branch Technical Position 11-5 and demonstrates compliance with regulatory limits. The dose consequence analysis evaluates a postulated event in which the GRWS fails. The analysis used in determining the radionuclide content of the effluents assumes that 1 percent of the operating fission product inventory in the core is released to the primary coolant. Table 11.3-8 tabulates the release source term. The dose consequences are calculated using the Radionuclide Transport and Removal and Dose (RADTRAD) code using the two-hour exclusion area boundary atmospheric dispersion factor from Table 2.0-1. Table 11.3-8 presents the resultant offsite doses.

L Item 11.3-3: An applicant that references the NuScale Power Plant US460 standard design will perform an analysis in accordance with Branch Technical Position 11-5 using the site-specific parameters.

cale US460 SDAA 11.3-5 Revision 1

The design of the ventilation systems for normal operation is in accordance with RG 1.140, and is described in Section 9.4.

3.5 Instrumentation and Controls The instruments that provide automated functions in the GRWS include the following:

3.5.1 Waste Gas Cooler Moisture Separator Level The waste gas cooler moisture separator level instrument monitors the water level in the drain tank and opens the tank's drain valve to route the water to the RWDS.

3.5.2 Hydrogen and Oxygen Gas Analyzers Section 11.3.2.1.6 describes the hydrogen and oxygen gas analyzers.

3.5.3 Fire Detectors Each of the charcoal beds has fire detectors to indicate the presence of a fire. If a fire is detected in a guard or decay bed, the GRWS waste gas inlet valve automatically closes and the nitrogen supply valve automatically opens to the associated charcoal bed.

3.5.4 Waste Gas Flow Instrument The waste gas flow instrument is downstream of the moisture separator and downstream of the decay beds in the outlet line. Nitrogen flow maintains maintain a minimum flow through the charcoal beds.

3.5.5 Moisture Instrument The waste gas stream contains a moisture level instrument at the outlet of the guard bed. If high moisture is detected, the waste gas inlet valve to the GRWS closes to stop the system flow.

3.5.6 Charcoal Bed Process Radiation Monitors The outlet of each of the two charcoal decay beds has process radiation monitors.

If high radiation is detected, the charcoal bed outlet valve closes.

3.5.7 Gaseous Radioactive Waste System Outlet Process Radiation Monitor The outlet of the GRWS also has a radiation monitor. If high radiation is detected, the GRWS outlet valve closes to stop system flow to the RWBVS.

cale US460 SDAA 11.3-6 Revision 1

Each of the charcoal bed cubicles has area airborne radiation detectors. If high radiation is detected, the waste gas inlet valve to the GRWS closes and the nitrogen purge valve opens.

3.6 Reference 11.3-1 NuScale Power, LLC, "Effluent Release (GALE Replacement)

Methodology and Results, TR-123242, Revision 1.

cale US460 SDAA 11.3-7 Revision 1

Parameter Nominal Value on delay 69 days (normal) ton delay 2.9 days (normal) amic adsorption coefficient (Kd) for xenon 1400 cm3/g amic adsorption coefficient (Kd) for krypton 60 cm3/g imum gas waste stream temperature 200 °F vated carbon operating temperature 50-105 °F flow rate 1.03 scfm (normal) rcoal particle size 0.132 inch cale US460 SDAA 11.3-8 Revision 1

Equipment / Parameter Description / Value Vapor Condenser Package Assembly ntity 2 ign pressure 150 psig ign temperature 250 °F gas design flow rate 3.5 scfm gas inlet temperature 200 °F chilled water inlet temperature 40.5 °F erial Stainless Steel 1.143 safety classification RW-IIc e for Assumed Radioactive Content Table 11.3-9 Charcoal Drying Heater ntity 1 e Electric 2.28 scfm mum Temperature Inlet -10 °F perature outlet 140 °F Charcoal Guard Bed ntity 1 e cylindrical pressure vessel inal volume 10 ft3 ign pressure 125 psig ign temperature 250 °F ign flow rate 3.5 scfm erial Stainless Steel 1.143 safety classification RW-IIa e for Assumed Radioactive Content Table 12.2-15 Charcoal Decay Bed Vessel ntity 2 e cylindrical pressure vessel inal volume 147.5 ft3/sec ign pressure 125 psig ign temperature 250 °F erial Stainless Steel 1.143 safety classification RW-IIa e for Assumed Radioactive Content Table 12.2-15 cale US460 SDAA 11.3-9 Revision 1

quipment Malfunction Results (Consequences) Mitigating or Alternate Action Item or Skid Failure Failure of the gas cooler causes If one of the condenser package assemblies fail, denser ineffective removal of moisture in the influent gas stream can be diverted to the kage the gas. other vapor condenser package. This allows embly processing to continue.

rcoal Drying Heater The purpose of the heater is to heat The charcoal decay bed skids may be aligned in ter Failure nitrogen used to periodically dry the series to improve the decontamination factor if charcoal in the charcoal guard bed the guard bed is saturated with moisture.

by allowing hot nitrogen to flow through the charcoal. There is no immediate impact to the GRW operation if the heater fails. The downstream charcoal decay bed efficiency may be lowered.

rcoal Guard Guard Bed There is only one charcoal guard Operation may continue by sending the gas to be Failure be in the system. If the guard bed treated in one of the two charcoal decay beds, fails, the fission gas removal which is located downstream of the bed.

efficiency may be lowered. Pressure differential transmitter monitors different pressures across the guard bed.

Detect moisture content of the gaseous stream through the moisture monitor located downstream of the guard bed.

The charcoal decay bed skids may be aligned in series to improve the decontamination factor if the guard bed is saturated with moisture.

rcoal Decay Decay Bed There is one set of redundant If one decay bed skid fails, the gas can be Skids Failure charcoal decay bed skids. Failure switched to the redundant decay bed skid for of one decay bed skid decreases continued operation.

removal efficiency of radioactive Radiation monitors downstream of the charcoal noble gases. decay beds alarm when high radiation level is detected in the effluent gas. A high radiation alarm also triggers automatic closure of the isolation valves to the RWBV. Operators should review indications and determine whether to isolate influent streams to the GRW as well.

sure Gas Leaks Waste gas is released to the RWB. Very small gas leaks in the GRW Charcoal Beds ndary room can be detected by the area airborne radiation monitors.

The system can be purged with nitrogen before repair of replacement of the leaking component.

gen Monitor Fail to Monitoring capability is lost in The redundant oxygen analyzer monitors the monitor detecting oxygen concentration. oxygen concentration downstream of the vapor condensers. A single oxygen analyzer is placed just before the discharge to RWBV.

The oxygen analyzers are set to alarm at 1% and 2%.

The high-high oxygen alarm with a set point of 2% isolates the input streams from CE and LRW.

Operators allow the continuous nitrogen flow to purge the system.

cale US460 SDAA 11.3-10 Revision 1

quipment Malfunction Results (Consequences) Mitigating or Alternate Action Item rogen Fail to Monitoring capability is lost in The redundant hydrogen analyzer monitors the itor monitor detecting hydrogen concentration. hydrogen concentration downstream of the vapor condensers.

Hydrogen analyzers are set to alarm at 2% and 4%.

iation Fail to There are two types of radiation In-line process radiation is monitored at each of itor monitor/Loss monitors: one is an area airborne the decay skid outlets and at the system of power monitor and the other is an in-line discharge providing redundancy for the process.

process radiation monitor. In both Therefore, failure of one does not impact cases, monitoring capability is lost operation of the other unless there is a loss of in detecting leakage to the GRW power.

rooms housing the equipment and Upon loss of power to the radiation monitors the system and any waste gas inlet and outlet valves are closed to isolate the released through the doors to the GRW.

RWB.

cale US460 SDAA 11.3-11 Revision 1

NuScale Effluent Source Term Model Assumption Value (1 NPM) Value (6 NPMs) asifier partition fractions:

ble gases 1 1 logens 0.5 0.5 ctor pool evaporation rate - 1300 lb/hour l evaporation partition fractions:

logens (except iodine) 0.01 0.01 ine 0.0005 0.0005

, Rb, particulates 0.005 0.005 ses and tritium 1 1 m generator partition coefficient 1 1

-efficiency particulate air filter particulate efficiency 0 0 ary coolant system leakrate 11.8 lb/day 70.6 lb/day ary coolant leak flashing fraction 0.4 0.4 ary coolant leak partition fractions:

logens 0.01 0.01

, Rb, particulates 0.005 0.005 ses and tritium 1 1 ondary coolant system steam leakrate 125 lb/day 750 lb/day denser air removal normalized iodine release rate 125 Ci/yr/µCi/gm 750 Ci/yr/µCi/gm tainment vessel design leakrate 0.2 weight%/day 0.2 weight%/day tainment depressurization time 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> cale US460 SDAA 11.3-12 Revision 1

Scale Final Safety Analysis Report Total Gaseous Condenser Effluent Primary Plant Secondary Air Concentration 10 CFR 20 Pool AOO Gas Coolant Exhaust Steam Removal Total TGB at Site Appendix B GRWS Evaporation Leakage Leaks Stack Total Leaks System Releases Boundary Limits Fraction of clide (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (µCi/ml) (µCi/ml) Limit m 6.5E-07 1.5E-07 8.8E-05 9.1E-03 9.1E-03 8.2E-07 4.2E-03 4.2E-03 4.2E-15 5.0E-05 8.5E-11 m 6.0E-05 - 3.7E-04 3.8E-02 3.8E-02 3.4E-06 1.8E-02 1.8E-02 1.8E-14 1.0E-07 1.8E-07 1.4E+02 - 1.8E-02 1.9E+00 1.4E+02 1.7E-04 8.9E-01 8.9E-01 4.6E-11 7.0E-07 6.6E-05 5.2E-17 - 2.0E-04 2.1E-02 2.1E-02 1.9E-06 9.7E-03 9.7E-03 9.7E-15 2.0E-08 4.8E-07 1.9E-07 - 5.9E-04 6.0E-02 6.1E-02 5.4E-06 2.8E-02 2.8E-02 2.8E-14 9.0E-09 3.1E-06

- - 1.3E-05 1.4E-03 1.4E-03 1.2E-07 6.4E-04 6.4E-04 6.4E-16 - -

31m 2.0E-01 3.6E-01 1.4E-03 1.5E-01 7.0E-01 1.3E-05 6.9E-02 6.9E-02 2.4E-13 2.0E-06 1.2E-07 33m 3.2E-07 4.2E-01 1.3E-03 1.3E-01 5.6E-01 1.2E-05 6.3E-02 6.3E-02 2.0E-13 6.0E-07 3.3E-07 3 8.4E-02 6.0E+00 9.5E-02 9.7E+00 1.6E+01 8.8E-04 4.6E+00 4.6E+00 6.5E-12 5.0E-07 1.3E-05 35m 7.3E-05 5.4E-02 1.3E-04 1.3E-02 6.7E-02 1.2E-06 6.1E-03 6.1E-03 2.3E-14 4.0E-08 5.8E-07 35 3.2E-05 3.0E-02 2.6E-03 2.7E-01 3.0E-01 2.4E-05 1.3E-01 1.3E-01 1.4E-13 7.0E-08 1.9E-06 37 - - 4.4E-05 4.5E-03 4.6E-03 4.1E-07 2.1E-03 2.1E-03 2.1E-15 - -

8 - - 1.5E-04 1.5E-02 1.6E-02 1.4E-06 7.2E-03 7.2E-03 7.2E-15 2.0E-08 3.6E-07 9.5E-09 6.4E-09 - 1.0E-06 1.0E-06 2.3E-08 5.7E-09 2.8E-08 3.3E-19 5.0E-09 6.6E-11 5.4E-08 1.1E-14 - 5.7E-06 5.8E-06 1.3E-07 3.2E-08 1.6E-07 1.9E-18 9.0E-08 2.1E-11 2.5E-08 - - 2.7E-06 2.7E-06 5.5E-08 1.4E-08 6.9E-08 8.7E-19 8.0E-08 1.1E-11 3.1E-09 - - 3.2E-07 3.2E-07 3.5E-09 8.7E-10 4.3E-09 1.0E-19 - -

Gaseous Waste Management System 1.6E-13 3.1E-13 - 1.6E-11 1.7E-11 3.7E-13 9.4E-14 4.7E-13 5.5E-24 4.0E-11 1.4E-13 7.7E-08 6.8E-09 - 8.2E-06 8.2E-06 1.8E-07 4.7E-08 2.3E-07 2.7E-18 3.0E-09 8.9E-10 2.0E-06 3.2E-04 - 2.1E-04 5.3E-04 4.7E-06 1.2E-06 5.9E-06 1.7E-16 2.0E-10 8.5E-07 9.1E-07 7.9E-07 - 9.6E-05 9.8E-05 2.1E-06 5.4E-07 2.7E-06 3.2E-17 2.0E-08 1.6E-09 3.0E-06 2.8E-05 - 3.2E-04 3.5E-04 7.2E-06 1.8E-06 9.0E-06 1.1E-16 1.0E-09 1.1E-07 5.4E-07 8.5E-26 - 5.7E-05 5.7E-05 1.2E-06 3.1E-07 1.5E-06 1.9E-17 6.0E-08 3.1E-10 1.9E-06 1.0E-08 - 2.0E-04 2.0E-04 4.5E-06 1.1E-06 5.6E-06 6.5E-17 6.0E-09 1.1E-08 6m - - - 1.2E-10 1.2E-10 1.3E-12 - 1.3E-12 3.9E-23 - -

6 - 8.3E-07 - 7.0E-07 1.5E-06 3.5E-08 - 3.5E-08 5.0E-19 1.0E-09 5.0E-10 8 - - - 1.2E-04 1.2E-04 5.0E-06 - 5.0E-06 4.0E-17 9.0E-08 4.4E-10 9 - - - 5.5E-06 5.5E-06 2.2E-07 - 2.2E-07 1.8E-18 2.0E-07 9.1E-12

Scale Final Safety Analysis Report Total Gaseous Condenser Effluent Primary Plant Secondary Air Concentration 10 CFR 20 Pool AOO Gas Coolant Exhaust Steam Removal Total TGB at Site Appendix B GRWS Evaporation Leakage Leaks Stack Total Leaks System Releases Boundary Limits Fraction of clide (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (µCi/ml) (µCi/ml) Limit 32 - 1.4E-08 - 1.4E-08 2.8E-08 7.0E-10 - 7.0E-10 9.1E-21 6.0E-09 1.5E-12 34 - 1.3E-04 - 1.0E-04 2.3E-04 5.0E-06 - 5.0E-06 7.4E-17 2.0E-10 3.7E-07 35m - 1.1E-26 - 8.4E-08 8.4E-08 3.9E-09 - 3.9E-09 2.8E-20 3.0E-07 9.3E-14 36 - 2.5E-05 - 2.2E-05 4.7E-05 1.1E-06 - 1.1E-06 1.5E-17 9.0E-10 1.7E-08 37 - 6.6E-05 - 5.2E-05 1.2E-04 2.6E-06 - 2.6E-06 3.8E-17 2.0E-10 1.9E-07 38 - - - 4.5E-05 4.5E-05 2.0E-06 - 2.0E-06 1.5E-17 8.0E-08 1.9E-10

- 6.9E-13 - 2.0E-12 2.7E-12 8.9E-14 - 8.9E-14 8.7E-25 5.0E-10 1.7E-15 7 - 5.8E-15 - 1.5E-14 2.1E-14 6.8E-16 - 6.8E-16 6.8E-27 9.0E-10 7.5E-18

- 3.4E-08 - 8.9E-08 1.2E-07 4.1E-09 - 4.1E-09 4.0E-20 2.0E-10 2.0E-10

- 5.4E-09 - 1.4E-08 1.9E-08 6.3E-10 - 6.3E-10 6.3E-21 6.0E-12 1.1E-09

- 3.2E-10 - 4.6E-08 4.7E-08 2.1E-09 - 2.1E-09 1.5E-20 5.0E-09 3.1E-12

- 4.6E-15 - 2.5E-08 2.5E-08 1.1E-09 - 1.1E-09 8.3E-21 9.0E-09 9.2E-13

- 3.2E-09 - 3.4E-09 6.6E-09 1.5E-10 - 1.5E-10 2.1E-21 9.0E-10 2.4E-12 m - 2.1E-10 - 2.5E-08 2.5E-08 1.1E-09 - 1.1E-09 8.3E-21 2.0E-07 4.1E-14

- 5.0E-09 - 1.3E-08 1.8E-08 5.9E-10 - 5.9E-10 5.9E-21 2.0E-10 2.9E-11

- 6.5E-13 - 2.1E-08 2.1E-08 9.5E-10 - 9.5E-10 7.0E-21 1.0E-08 7.0E-13

- 8.6E-11 - 1.0E-08 1.0E-08 4.5E-10 - 4.5E-10 3.3E-21 3.0E-09 1.1E-12 Gaseous Waste Management System

- 5.7E-10 - 1.5E-08 1.5E-08 6.7E-10 - 6.7E-10 5.1E-21 2.0E-09 2.5E-12 5 - 3.9E-05 - 3.7E-08 3.9E-05 1.7E-09 - 1.7E-09 1.2E-17 2.0E-09 6.2E-09 9 - 5.8E-06 - 2.7E-05 3.2E-05 1.2E-06 - 1.2E-06 1.1E-17 2.0E-09 5.3E-09 01 - - - 1.0E-06 1.0E-06 3.8E-08 - 3.8E-08 3.3E-19 2.0E-07 1.7E-12 m - 5.6E-06 - 2.5E-05 3.0E-05 1.1E-06 - 1.1E-06 9.9E-18 2.0E-07 4.9E-11

- 2.0E-10 - 5.0E-10 7.0E-10 2.3E-11 - 2.3E-11 2.3E-22 8.0E-09 2.8E-14 03 - 9.5E-09 - 2.5E-08 3.5E-08 1.2E-09 - 1.2E-09 1.1E-20 9.0E-10 1.3E-11 05 - 5.3E-13 - 8.4E-09 8.4E-09 3.8E-10 - 3.8E-10 2.8E-21 2.0E-08 1.4E-13 06 - 6.2E-09 - 1.6E-08 2.2E-08 7.2E-10 - 7.2E-10 7.2E-21 2.0E-11 3.6E-10 03m - 9.4E-09 - 2.5E-08 3.4E-08 1.1E-09 - 1.1E-09 1.1E-20 2.0E-06 5.6E-15 05 - 2.4E-09 - 1.7E-08 2.0E-08 7.8E-10 - 7.8E-10 6.4E-21 8.0E-09 8.1E-13

Scale Final Safety Analysis Report Total Gaseous Condenser Effluent Primary Plant Secondary Air Concentration 10 CFR 20 Pool AOO Gas Coolant Exhaust Steam Removal Total TGB at Site Appendix B GRWS Evaporation Leakage Leaks Stack Total Leaks System Releases Boundary Limits Fraction of clide (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (µCi/ml) (µCi/ml) Limit 06 - 6.2E-09 - 1.6E-08 2.2E-08 9.7E-11 - 9.7E-11 7.0E-21 - -

10 - 4.0E-05 - 1.1E-07 4.0E-05 5.7E-10 - 5.7E-10 1.3E-17 - -

24 - 1.4E-11 - 3.8E-11 5.2E-11 1.7E-12 - 1.7E-12 1.7E-23 3.0E-10 5.7E-14 25 - 1.1E-10 - 2.8E-10 3.9E-10 1.3E-11 - 1.3E-11 1.3E-22 7.0E-10 1.8E-13 27 - 3.7E-10 - 1.4E-09 1.8E-09 6.5E-11 - 6.5E-11 5.9E-22 1.0E-09 5.9E-13 29 - 1.0E-13 - 1.8E-09 1.8E-09 7.9E-11 - 7.9E-11 5.8E-22 1.0E-08 5.8E-14 5m - 1.6E-08 - 4.1E-08 5.6E-08 1.9E-09 - 1.9E-09 1.8E-20 1.0E-09 1.8E-11 7m - 6.0E-08 - 1.6E-07 2.2E-07 7.1E-09 - 7.1E-09 7.0E-20 4.0E-10 1.8E-10 7 - 6.2E-08 - 6.2E-07 6.8E-07 2.8E-08 - 2.8E-08 2.2E-19 2.0E-08 1.1E-11 9m - 1.7E-07 - 4.5E-07 6.1E-07 2.0E-08 - 2.0E-08 2.0E-19 3.0E-10 6.7E-10 9 - 1.0E-07 - 6.3E-07 7.4E-07 2.8E-08 - 2.8E-08 2.4E-19 9.0E-08 2.7E-12 1m - 1.6E-07 - 1.5E-06 1.6E-06 6.6E-08 - 6.6E-08 5.3E-19 1.0E-09 5.3E-10 1 - 3.5E-08 - 7.2E-07 7.6E-07 2.9E-08 - 2.9E-08 2.5E-19 2.0E-08 1.2E-11 2 - 2.5E-06 - 1.1E-05 1.3E-05 4.8E-07 - 4.8E-07 4.3E-18 9.0E-10 4.8E-09 3m - 2.4E-25 - 9.2E-07 9.2E-07 3.9E-08 - 3.9E-08 3.0E-19 2.0E-08 1.5E-11 4 - - - 1.3E-06 1.3E-06 5.5E-08 - 5.5E-08 4.3E-19 7.0E-08 6.1E-12 37m - 6.2E-05 - 4.8E-05 1.1E-04 9.7E-07 - 9.7E-07 3.5E-17 - -

Gaseous Waste Management System 39 - 6.9E-21 - 2.4E-08 2.4E-08 1.1E-09 - 1.1E-09 8.0E-21 4.0E-08 2.0E-13 40 - 4.5E-08 - 1.3E-07 1.8E-07 5.9E-09 - 5.9E-09 5.7E-20 2.0E-09 2.9E-11 0 - 3.5E-08 - 3.8E-08 7.3E-08 1.7E-09 - 1.7E-09 2.4E-20 2.0E-09 1.2E-11 1 - 1.5E-13 - 7.5E-09 7.5E-09 3.4E-10 - 3.4E-10 2.5E-21 1.0E-08 2.5E-13 2 - 1.2E-20 - 3.6E-09 3.6E-09 1.6E-10 - 1.6E-10 1.2E-21 3.0E-08 3.9E-14 41 - 7.5E-09 - 2.0E-08 2.8E-08 9.1E-10 - 9.1E-10 9.0E-21 8.0E-10 1.1E-11 43 - 1.8E-09 - 1.5E-08 1.7E-08 6.9E-10 - 6.9E-10 5.6E-21 2.0E-09 2.8E-12 44 - 6.6E-09 - 1.7E-08 2.4E-08 7.7E-10 - 7.7E-10 7.7E-21 2.0E-11 3.9E-10 3 - 6.5E-09 - 1.8E-08 2.4E-08 8.1E-10 - 8.1E-10 7.9E-21 9.0E-10 8.8E-12 4 - 6.5E-09 - 1.7E-08 2.3E-08 6.5E-10 - 6.5E-10 7.6E-21 2.0E-07 3.8E-14 39 - 6.3E-08 - 3.2E-07 3.8E-07 1.5E-08 - 1.5E-08 1.3E-19 3.0E-09 4.2E-11

Scale Final Safety Analysis Report Total Gaseous Condenser Effluent Primary Plant Secondary Air Concentration 10 CFR 20 Pool AOO Gas Coolant Exhaust Steam Removal Total TGB at Site Appendix B GRWS Evaporation Leakage Leaks Stack Total Leaks System Releases Boundary Limits Fraction of clide (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) (µCi/ml) (µCi/ml) Limit 4 - 9.4E-06 - 3.2E-04 3.3E-04 1.5E-05 - 1.5E-05 1.1E-16 7.0E-09 1.6E-08

- 6.7E-03 - 1.8E-05 6.7E-03 8.2E-07 - 8.2E-07 2.1E-15 3.0E-08 7.1E-08 4 - 3.6E-03 - 9.3E-06 3.6E-03 4.2E-07 - 4.2E-07 1.1E-15 1.0E-09 1.1E-06

- 2.7E-03 - 7.0E-06 2.7E-03 3.2E-07 - 3.2E-07 8.6E-16 3.0E-09 2.9E-07

- 6.6E-04 - 1.7E-06 6.6E-04 8.0E-08 - 8.0E-08 2.1E-16 5.0E-10 4.2E-07 8 - 1.0E-01 - 2.7E-05 1.0E-01 1.2E-06 - 1.2E-06 3.2E-14 1.0E-09 3.2E-05 0 - 1.2E-03 - 3.1E-06 1.2E-03 1.4E-07 - 1.4E-07 3.8E-16 5.0E-11 7.6E-06

- 6.0E-04 - 1.5E-06 6.0E-04 7.0E-08 - 7.0E-08 1.9E-16 2.0E-09 9.5E-08

- 1.2E-03 - 3.0E-06 1.2E-03 1.3E-07 - 1.3E-07 3.7E-16 4.0E-10 9.1E-07

- 8.7E-04 - 2.3E-06 8.7E-04 1.0E-07 - 1.0E-07 2.7E-16 4.0E-10 6.9E-07 10m - 2.9E-03 - 7.5E-06 2.9E-03 3.4E-07 - 3.4E-07 9.3E-16 1.0E-10 9.3E-06 7 - 1.2E-03 - 1.6E-05 1.3E-03 7.4E-07 - 7.4E-07 4.0E-16 1.0E-08 4.0E-08

- 6.6E+02 - 6.3E+00 6.7E+02 7.4E+00 - 7.4E+00 2.1E-10 1.0E-07 2.1E-03 2.3E-01 5.4E-03 - 1.2E-03 2.4E-01 2.8E-07 - 2.8E-07 7.5E-14 3.0E-09 2.5E-05 2.3E+00 - 1.6E-02 1.6E+00 4.0E+00 1.5E-04 7.7E-01 7.7E-01 1.5E-12 1.0E-08 1.5E-04 l 1.5E+02 6.7E+02 1.4E-01 2.0E+01 8.4E+02 7.4E+00 6.5E+00 1.4E+01 2.7E-10 5.8E-05 2.5E-03

- The X/Q used to calculate the site boundary concentrations is provided in Table 11.3-6 Gaseous Waste Management System

Parameter Value tine release /Q (undepleted/no decay) Table 2.0-1 tine release D/Q Table 2.0-1 animal Greater of goat or cow point of plant life 20 yrs tion of year that leafy vegetables are grown 1.0 tion of year that milk cows are in pasture 1.0 tion of the maximum individuals vegetable intake that is from his own garden 0.76 tion of milk-cow feed intake that is from pasture while on pasture 1.0 rage absolute humidity over the growing season 8.0 gram/m3 tion of year that beef cattle are in pasture 1.0 tion of beef cattle feed intake that is from pasture while the cattle are on pasture 1.0 rce term Table 11.3-5 cale US460 SDAA 11.3-17 Revision 1

Type of Dose Dose Estimate Beta Dose Air (mrad/yr) 0.10 Gamma Dose Air (mrad/yr) 0.02 HWAY T.BODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG SKIN e 1.2E-02 1.2E-02 1.2E-02 1.2E-02 1.2E-02 1.2E-02 1.3E-02 8.2E-02 und 2.7E-01 2.7E-01 2.7E-01 2.7E-01 2.7E-01 2.7E-01 2.7E-01 3.2E-01 ETABLE ADULT 3.4E-01 6.6E-01 1.0E-01 3.3E-01 3.0E-01 3.9E-01 3.0E-01 2.9E-01 TEEN 4.1E-01 7.1E-01 1.6E-01 4.0E-01 3.6E-01 4.6E-01 3.5E-01 3.4E-01 CHILD 6.7E-01 7.8E-01 3.9E-01 6.4E-01 5.7E-01 7.7E-01 5.6E-01 5.5E-01 T

ADULT 6.0E-02 1.8E-01 3.3E-02 5.8E-02 4.8E-02 4.9E-02 4.6E-02 4.5E-02 TEEN 4.0E-02 1.0E-01 2.7E-02 3.8E-02 3.0E-02 3.1E-02 2.9E-02 2.8E-02 CHILD 5.5E-02 7.3E-02 5.1E-02 4.9E-02 4.0E-02 4.2E-02 3.8E-02 3.7E-02 W MILK ADULT 1.2E-01 3.1E-01 5.0E-02 1.3E-01 1.1E-01 2.2E-01 1.0E-01 1.0E-01 TEEN 1.6E-01 3.7E-01 8.9E-02 1.8E-01 1.6E-01 3.2E-01 1.4E-01 1.3E-01 CHILD 2.6E-01 3.8E-01 2.2E-01 3.0E-01 2.6E-01 5.8E-01 2.2E-01 2.2E-01 INFANT 4.0E-01 5.3E-01 3.7E-01 4.9E-01 4.0E-01 1.2E+00 3.5E-01 3.5E-01 AT MILK ADULT 2.3E-01 2.2E-01 4.7E-02 2.3E-01 2.1E-01 3.4E-01 2.0E-01 2.0E-01 TEEN 2.9E-01 2.9E-01 8.5E-02 3.2E-01 2.8E-01 4.8E-01 2.7E-01 2.6E-01 CHILD 4.5E-01 4.4E-01 2.0E-01 5.2E-01 4.6E-01 8.6E-01 4.3E-01 4.2E-01 INFANT 6.8E-01 6.8E-01 3.7E-01 8.4E-01 7.1E-01 1.7E+00 6.7E-01 6.5E-01 ALATION ADULT 1.5E-01 1.6E-01 2.1E-04 1.5E-01 1.5E-01 1.6E-01 1.9E-01 1.5E-01 TEEN 1.6E-01 1.6E-01 2.8E-04 1.6E-01 1.6E-01 1.6E-01 2.1E-01 1.6E-01 CHILD 1.4E-01 1.4E-01 3.8E-04 1.4E-01 1.4E-01 1.4E-01 1.8E-01 1.4E-01 INFANT 7.9E-02 7.9E-02 1.9E-04 7.9E-02 7.9E-02 8.2E-02 1.1E-01 7.9E-02 AL ADULT 2.8E-01 1.3E+00 1.9E-01 7.8E-01 7.2E-01 9.3E-01 7.4E-01 4.0E-01 TEEN 2.8E-01 1.3E+00 2.8E-01 9.2E-01 8.2E-01 1.1E+00 8.5E-01 4.0E-01 CHILD 2.8E-01 1.4E+00 6.6E-01 1.4E+00 1.2E+00 1.8E+00 1.2E+00 4.0E-01 INFANT 2.8E-01 7.6E-01 3.7E-01 9.2E-01 7.9E-01 1.8E+00 7.8E-01 4.0E-01 cale US460 SDAA 11.3-18 Revision 1

System Failure Parameter Value ase Source Term:

I-131 4.1E-04 Ci I-132 1.9E-04 Ci I-133 6.2E-04 Ci I-134 1.1E-04 Ci I-135 3.9E-04 Ci Xe-133 5.3E-02 Ci Xe-135 9.7E+00 Ci Kr-85m 2.9E-02 Ci Kr-85 8.4E-02 Ci Kr-87 1.4E+01 Ci Kr-88 3.8E-01 Ci ersion factor (0-2 hour exclusion area boundary) Table 2.0-1 ite dose consequence < 10 mrem wable dose limit 100 mrem cale US460 SDAA 11.3-19 Revision 1

Isotope Activity (Ci/cm3)

Kr83m 1.2E-10 Kr85m 5.1E-10 Kr85 9.2E-08 Kr87 2.8E-10 Kr88 8.0E-10 Kr89 1.8E-11 Xe131m 2.1E-09 Xe133m 1.8E-09 Xe133 1.3E-07 Xe135m 1.7E-10 Xe135 3.6E-09 Xe137 6.0E-11 Xe138 2.1E-10 Br82 1.8E-14 Br83 1.0E-13 Br84 4.8E-14 Br85 5.8E-15 I129 3.0E-19 I130 1.5E-13 I131 3.8E-12 I132 1.7E-12 I133 5.7E-12 I134 1.0E-12 I135 3.6E-12 C14 1.6E-11 Ar41 1.3E-08 cale US460 SDAA 11.3-20 Revision 1

Scale Final Safety Analysis Report SSC (Note 1) Location SSC Augmented Design Quality Group/Safety Seismic Classification Classification Requirements Classification (Ref RG (Ref. RG 1.29 or RG (A1, A2, B1, B2) (Note 2) 1.26 or RG 1.143) 1.143) (Note 4)

(Note 3)

GRWS, Gaseous Radioactive Waste System omponents (except those listed below): RWB B2 RG 1.143 RW-IIc III arcoal decay bed skid & associated valves RWB B2 RG 1.143 RW-Ila RW-Ila xcept inlet & outlet root valves) arcoal guard bed & associated valves trument root valves (RIT-1021A/B) trumentation RWB B2 None N/A III s sampler 1: Acronyms used in this table are listed in Table 1.1-1 2: Additional augmented design requirements, such as the application of a Quality Group, Radwaste safety, or seismic classification, to nonsafety-related SSC are reflected in the columns Quality Group / Safety Classification and Seismic Classification, where applicable. Environmental Qualifications of SSC are identified in Table 3.11-1.

3: Section 3.2.2.1 through Section 3.2.2.4 provides the applicable codes and standards for each RG 1.26 Quality Group designation (A, B, C, and D). A Quality Group classification per RG 1.26 is not applicable to supports or instrumentation. Section 3.2.1.4 provides a description of RG 1.143 classification for RW-IIa, RW-IIb, and RW-IIc.

4: Where SSC (or portions thereof) as determined in the as-built plant that are identified as Seismic Category III in this table could, as the result of a seismic event, adversely affect Seismic Category I SSC or result in incapacitating injury to occupants of the control room, they are categorized as Seismic Category II consistent with Section 3.2.1.2 and analyzed as described in Section 3.7.3.8.

Gaseous Waste Management System

NuScale Final Safety Analysis Report Gaseous Waste Management System Figure 11.3-1: Gaseous Radioactive Waste System Diagram GRWS CHARCOAL DRYING HEATER FROM NDS PRV-0053 FROM CHILLED WATER SUPPLY FROM CHILLED WATER SUPPLY GRWS GAS SAMPLER CHARCOAL GUARD BED VAPOR CONDENSER VAPOR CONDENSER PACKAGE ASSEMBLY A PACKAGE ASSEMBLY B GRWS CHARCOAL GRWS CHARCOAL TO RWBVS DECAY BED SKID A DECAY BED SKID B PRV-0021 TO CHILLED WATER RETURN TO CHILLED WATER RETURN RXB RWB TO RWDS TO RWDS FROM LRWS DEGASIFIER FROM CES CONDENSER FROM CES MODULES 02-06 NuScale US460 SDAA 11.3-22 Revision 1

The solid waste management system is called the solid radioactive waste system (SRWS). The SRWS is designed to process both wet solid waste (WSW) and dry solid waste (DSW) from various plant systems produced during normal operation and anticipated operational occurrences, including startup, shutdown, and refueling operations. The Radioactive Waste Building (RWB) has adequate space for onsite storage for various solid waste containers plus space for mobile processing equipment.

The SRWS includes the WSW system, DSW system, mixed waste system, and an onsite storage area.

The design basis source term identified in Section 11.1 forms the basis for the shielding design. The shield wall thickness evaluation assumes that the spent filters and spent resins fully loaded using the design basis source term. Section 12.3 discusses additional details on the shielding design.

The wet and dry radioactive solid waste packaged for offsite shipment and disposal complies with the requirements of 10 CFR 61.55, 10 CFR 61.56, 10 CFR 71 and 49 CFR 171-180, as applicable.

Onsite storage allows for radioactive decay with adequate storage in case of processing, maintenance or transportation delays. Onsite storage is adequate to hold solid waste for at least 30 days in accordance with ANSI/ANS-55.1-1992 (Reference 11.4-1) and BTP 11-3. The SRWS meets the design recommendations of BTP 11-3.

The SRWS and associated handling areas have area radiation monitoring equipment to detect excessive radiation or airborne levels and initiate appropriate alarms and procedural actions to maintain radiation exposure as low as reasonably achievable (ALARA). Section 12.3 provides additional information on area radiation monitors.

4.1 System Description The SRWS is a nonsafety-related system, serves no safety-related functions, and is not risk-significant. Table 11.4-5 identifies SSC classifications for the SRWS. The SRWS is designed to

  • collect, process, sample, package, and store WSW generated from the chemical and volume control system (CVCS), pool cooling and cleanup system, and liquid radioactive waste system (LRWS), using both permanently installed and mobile equipment in the SRWS.
  • collect, segregate, sample, package, and store compactible and non-compactible DSW.
  • collect, sample, segregate, package, and ship mixed and oily wastes.
  • provide sufficient storage space for packaged solid wastes.
  • process and package waste into disposal containers that are approved by the Department of Transportation and are acceptable to licensed waste disposal facilities for offsite shipment and burial.

cale US460 SDAA 11.4-1 Revision 1

  • transfer liquid wastes to the RWDS or LRWS.

The SRWS design handles three types of generated wastes: WSWs, DSWs, and miscellaneous wastes.

The boundaries of the SRWS begin at the connection to a particular waste stream source and end at the packaged waste container offsite shipment. For WSW, these connections usually involve flanged joints, and boundary valves at the system inlets.

For DSW, the boundaries are not always physical because much of DSW is collected from a variety of locations and transported through corridors to the solid radioactive waste sorting area.

For spent resins and granular activated charcoal, the SRWS starts downstream of the boundary valve from each demineralizer and carbon bed. Operators sluice spent resin into the SRSTs or PSTs for decay, and to waste containers.

For spent cartridge filters, the SRWS starts at the filter extraction point. Operators remove the spent filter from the filter housing and place it in a shielded spent filter transfer cask.

4.1.1 Dry Solid Waste Dry solid waste includes heating ventilation and air conditioning filters, tools and equipment, used personnel protective equipment, rags, paper, wood and miscellaneous cleaning supplies. Figure 11.4-1 summarizes the DSW handling and storage operation.

During some anticipated operational occurrences, such as refueling, the rate of DSW generation is higher than during normal operations. Major equipment items, such as core components and containment vessel components, are not processed in the SRWS.

4.1.2 Wet Solid Waste The WSW processing system receives and processes three major waste streams:

  • radioactive spent resin and spent charcoal
  • spent cartridge filters
  • filter membranes and reverse osmosis The WSW is homogenized, sampled, and analyzed to classify the waste in accordance with 10 CFR 61. Operators transfer spent resin and spent charcoal to high integrity containers (HICs) that are connected to a dewatering system located inside a confined enclosure.

Operators cap and seal containers after dewatering, and survey and decontaminate the containers, as necessary, to meet 49 CFR 173 requirements.

cale US460 SDAA 11.4-2 Revision 1

or other suitable containers and transfer the containers to the SRWS area for processing and storage.

In accordance with BTP 11-3, components and piping that contain slurries have flushing capabilities via the LRW clean-in-place skid or directly from the demineralized water break tank. The spent resin storage and PSTs are ASME Section VIII tanks that can use compressed service air to pressurize the tanks and pneumatically transport resin to a HIC. The associated pressure relief valves on the spent resin storage and PSTs are vented to the tank's cubicle, which are vented to the RWBVS. The hooded vents on the SRWS PST and SRST consist of tank vent piping that terminates below a vent hood and directs air into the RWBVS. The vent piping exiting the storage tanks contains an internal screen designed to prevent solids (i.e., resin) from escaping. An air gap between the tank vent piping and the vent hood minimizes contamination from entering the RWBVS. Liquid overflow flows out of a vent pipe into shielded cubicles lined with stainless steel.

Figure 11.4-2a and Figure 11.4-2b are process flow diagrams of the spent resin handling system.

To avoid the generation of explosive gas mixtures and exothermic reactions, the upstream systems (LRWS, pool cooling and cleanup system, CVCS) that transfer resins to the SRST or phase separator tank (PST) do not use chemicals (e.g.,

nitrates, nitrites) that can generate exothermic reactions with resins.

The main source of oily waste is expected to come from floor drains. Operators direct the oil to the SRWS from the LRWS oil separators and manually collect it in drums. The drums of contaminated oil are sent to an offsite treatment facility.

4.1.3 Mixed Waste Handling Mixed waste is a combination of radioactive waste mixed with Resource Conservation and Recovery Act-listed hazardous waste as defined in 40 CFR 261 Subpart D. The generation of mixed waste volume is expected to be low. Mixed waste can only be disposed of in a permitted mixed waste disposal facility.

Operators collect mixed waste near the source and transfer in drums to a permitted facility.

4.1.4 Packaging, Storage, and Shipping The Process Control Program (PCP) classifies waste as Class A, Class B, Class C, or greater than Class C in accordance with 10 CFR 61.55 and 10 CFR 61.56.

Table 11.4-2 and Table 11.4-3 provide the expected annual volumes of solid waste and shipment offsite estimates. The packaging and shipment of radioactive solid waste for disposal complies with 10 CFR 20, Appendix G, 10 CFR 61.56, and 49 CFR 173, Subpart I.

cale US460 SDAA 11.4-3 Revision 1

on the upper level for Class A waste. At the expected waste generation rates, there is storage capacity for at least 30 days.

The design and construction of SRWS components are in compliance with the codes and standards provided in RG 1.143. Each component is classified as RW-IIa, RW-IIb or RW-IIc based on the radionuclide content compared against the A1 and A2 values tabulated in 10 CFR 71, Appendix A. The safety classification for the SRWS components applies to components, up to and including the nearest isolation device. Table 11.4-1 provides design parameters for each of the major components.

4.1.4.1 Piping and Valves The SRWS piping material is stainless steel and is butt-welded to minimize crud traps. Backing rings are not allowed in SRWS piping. Slurry transport lines are sized to maintain a flow velocity to prevent the slurry from settling and utilize bends of five pipe diameter radius. Slurry lines are also sloped to promote complete drainage and are connected to the clean-in-place skid and directly to the demineralized water break tank to allow flushing and cleaning of SRWS piping and components after batch operations. Piping is also arranged to minimize tees, pipe branches, and dead legs. The SRWS valves are stainless steel, remote air-operated valves. Valves in slurry transfer lines are full-ported ball valves and liquid process valves are diaphragm valves.

4.1.4.2 Dewatering System The dewatering system is a skid-based, vendor-supplied package that removes free-standing water from waste packages to meet transportation and disposal requirements. The fillhead portion of the dewatering system includes an exhaust vent with high-efficiency particulate air filtration routed to the Radioactive Waste Building HVAC system (RWBVS) to control airborne contamination. Liquid removed by the dewatering system is routed to the LRWS low-conductivity waste collection tank. The dewatering system and associated connections to permanent plant equipment, including non-contaminated utilities, complies with IE Bulletin 80-10, Regulatory Guide 1.143, ANSI/ANS-55.1-1992 (Reference 11.4-1), and ANSI/ANS-40.37-2009 (Reference 11.4-2).

4.1.5 Effluent Controls The SRWS does not release effluents directly to the environment. Liquids removed from solid waste processing are transferred to the LRWS for further processing.

During the operation of the SRWS, such as processing and packaging solid waste, the expelled air is captured by the RWBVS to prevent unmonitored contamination being released to the environment.

cale US460 SDAA 11.4-4 Revision 1

Because the SRWS does not release effluents to the environment, a cost-benefit analysis is not performed separately from the evaluations in Section 11.2 and Section 11.3.

4.1.7 Mobile or Temporary Equipment The design of SRWS includes modular equipment (e.g., spent resin dewatering system) and options for additional mobile equipment (e.g., shredders, laundry unit). The purpose of modular and mobile equipment is to provide ease of equipment replacement due to either advances in treatment technologies or equipment problems.

4.2 Radioactive Effluent Releases The SRWS sends liquid and gaseous effluents to the LRWS and RWBVS, respectively. As a result, other than solid waste shipments offsite, the SRWS does not release effluents directly to the environment. The contributions to the offsite dose consequences from SRWS are included in the evaluations for LRW and gaseous radioactive waste systems in Section 11.2 and Section 11.3.

The SRWS design complies with the requirements of 10 CFR 20.1406. Section 12.3 discusses the SRWS design features to prevent the spread of contamination, facilitate decommissioning, and reduce the generation of radioactive waste.

The PCP follows the guidance of Nuclear Energy Institute 07-10A (Reference 11.4-3).

The PCP describes the administrative and operational controls used for the solidification of liquid or WSW and the dewatering of WSW.

4.3 Malfunction Analysis To demonstrate the design's resistance to failures, a malfunction analysis is performed. Table 11.4-4 summarizes this malfunction analysis.

4.4 Testing and Inspection Requirements The SRWS is tested during plant pre-operations to ensure operation of components and processes as discussed in Section 14.2. During plant operations, the periodic testing and inspection requirements of RG 1.143 are performed to support continued proper operation of components.

4.5 References 11.4-1 American National Standards Institute/American Nuclear Society, "Solid Radioactive Waste Processing System for Light-Water-Cooled Reactor Plants," ANSI/ANS-55.1-1992, LaGrange Park, IL.

cale US460 SDAA 11.4-5 Revision 1

ANSI/ANS-40.37-2009, LaGrange Park, IL.

11.4-3 Nuclear Energy Institute, "Generic FSAR Template Guidance for Process Control Program," NEI 07-10A, Revision 0, March 2009.

cale US460 SDAA 11.4-6 Revision 1

Scale Final Safety Analysis Report omponent (Quantity) RG 1.143 Standards Type Capacity Design Design Material Table for Safety Pressure Temperature Assumed Classification (psig) (°F) Radioactive Content nt resin storage tank (2) RW-IIa ASME BVPC Vertical Conical Austenitic 12.2-18 10,000 gal 175 155 Section VIII Stainless Steel T transfer pump (2) RW-IIc API-685 Sealless, centrifugal 75 gpm & Austenitic -

238 180 200 gpm Stainless Steel se separator tank (2) RW-IIa ASME BVPC Vertical Conical Austenitic 12.2-18 12,500 gal 175 155 Section VIII Stainless Steel transfer pump (2) RW-IIc API-685 Sealless, centrifugal 75 gpm 238 180 Stainless Steel -

atering skid (1) RW-IIc ANS-55.1 - 35 gpm 238 180 Stainless Steel -

pactor (1) - ANS-55.1 - 40 ft3 238 130 Stainless Steel -

Solid Waste Management System

Waste Sources and Waste Volume Container Container No. of Containers ssification Classification Generated Type Volume (rounded off)

(A or B/C) (ft3/yr) (ft3)

Class A Filters 194 B-25 box 90 3 Class A PPE/rags 2500 B-25 box 90 28 Class A Tools 9 drum 7.4 2 tal Class A 2700 Class B/C Failed equipment 14 drum 7.4 2 cale US460 SDAA 11.4-8 Revision 1

Waste Sources Volume Container Container No. of Containers ssification Generated Type Volume (rounded off)

(ft3/yr) (ft3)

Class B/C Resin 320 HIC 120 3 Class B/C Cartridge Filters 26 HIC 120 1 Class B/C Membrane Filters 4 Drum 7.4 1 al Class B/C 350 Class A Resin/Activated Charcoal 940 HIC 120 9 Class A Filters 4 Drum 7.4 1 Class A self-contained filter 24 IP-1 24 1 tal Class A 970 cale US460 SDAA 11.4-9 Revision 1

uipment Item Malfunction Results (Consequences) Mitigating or Alternate Actions nt Resin Tank failure There are two Spent Resin Storage If one of the tanks fail, the other tank age Tank Tanks to collect Class B/C spent can receive the PCW and CVC resins.

resins from the PCW and the CVC ion Alternate action includes direct exchangers. The PCW/CVC sluicing to the Dewatering Skid.

demineralizers would not be able to send spent resins to the Spent Resin Storage Tanks for processing.

nt Resin Pump failure There are two Spent Resin Transfer The tank transfer line is sfer Pump Pumps with one pump dedicated to cross-connected to the pump suction, each Spent Resin Storage Tank. The allowing the process to continue using consequence would be that liquid the redundant pump when one pump waste could not be transferred to the fails1.

LRW collection tanks and the pumps would not be able to transfer spent resin to the Spent Resin Storage Tanks or HICs.

se Separator Tank failure There are two Phase Separator Tanks If one tank fails, the other tank would k to collect Class A spent media. The continue to receive spent media.

LRW processing equipment or CVC Alternate action includes direct auxiliary ion exchangers would not be sluicing to the Dewatering Skid.

able to transfer spent media to the phase separator tanks or the HIC.

se Separator Pump failure There are two pumps, with one pump The separator transfer line is cross sfer Pump dedicated to each tank. The connected to the pump suction. The consequence would be that liquid other pump allows processing to waste could not be transferred to the continue when one pump fails1.

LRW collection tanks and the pumps would not be able to be used to transfer spent resin to the Phase Separator Tank or HICs.

atering Skid Skid component The major components for the No impact on collection of WSW. The failure Dewatering Skid are not equipped Phase Separator Tanks or the Spent with standby units. Excess liquid Resin Storage Tanks can store the waste from the HIC cannot be waste until dewatering skid extracted using the dewatering pump. components are repaired or replaced.

If the level control valve on the The HICs are equipped with a camera dewatering skid fails, the HIC may be on the fill head to monitor the HIC overflowed. level. The video monitors external leaks associated with the HIC.

Integrity Container is HICs are transported between the fill The grapple assembly has limit tainer dropped during station, storage area, and truck bay switches to ensure all the legs are transportation area. Dropping a HIC can cause local engaged prior to lifting the HIC. In contamination. The floor drains in the addition, the crane has its own safety area collect the liquid; however, the brake system to ensure the HIC is not solid portion needs to be removed by dropped during the power failure.

a shop vacuum.

nt Resin Service Air Supply The tanks cannot be pressurized to If one service air compressor fails, the age Tanks and perform pneumatic sluicing. backup service air compressor is used se Separator The Spent Resin Storage Tanks and to pressurize the tanks to complete ks Phase Separator Tanks would not be the sluicing2. If service air fails during able to send spent resins to the HICs the resin transfer, the lines are flushed for shipping offsite. to the HIC and system is restored to standby position.

cale US460 SDAA 11.4-10 Revision 1

uipment Item Malfunction Results (Consequences) Mitigating or Alternate Actions atering Skid Hose ruptures or The dewatering room is The video in the room shows the es flange failure contaminated, and resin slurry enters hoses connected to the HIC and the drainage system. dewatering operation. Operation is stopped manually, and affected hose is replaced after decontaminating the area.

s:

umps are provided with drain connections to the RWD to prevent the spread of contamination from leaks or from pairs.

ervice air compressors are part of the service air system (SAS). Section 9.3 contains information on the SAS.

cale US460 SDAA 11.4-11 Revision 1

Scale Final Safety Analysis Report SSC (Note 1) Location SSC Augmented Design Quality Group/Safety Seismic Classification Classification Requirements Classification (Ref RG (Ref. RG 1.29 or RG (A1, A2, B1, B2) (Note 2) 1.26 or RG 1.143) 1.143) (Note 4)

(Note 3)

SRWS, Solid Radioactive Waste System omponents (except those listed below) RWB/RXB B2 RG 1.143 RW-IIc III se separator tank (including strainers and RWB B2 RG 1.143 RW-IIa RW-IIa es) nt resin storage tank (including strainers and es) umentation RWB B2 None N/A III 1: Acronyms used in this table are listed in Table 1.1-1 2: Additional augmented design requirements, such as the application of a Quality Group, Radwaste safety, or seismic classification, to nonsafety-related SSC are reflected in the columns Quality Group / Safety Classification and Seismic Classification, where applicable. Environmental Qualifications of SSC are identified in Table 3.11-1.

3: Section 3.2.2.1 through Section 3.2.2.4 provides the applicable codes and standards for each RG 1.26 Quality Group designation (A, B, C, and D). A Quality Group classification per RG 1.26 is not applicable to supports or instrumentation. Section 3.2.1.4 provides a description of RG 1.143 classification for RW-IIa, RW-IIb, and RW-IIc.

4: Where SSC (or portions thereof) as determined in the as-built plant that are identified as Seismic Category III in this table could, as the result of a seismic event, adversely affect Seismic Category I SSC or result in incapacitating injury to occupants of the control room, they are categorized as Seismic Category II consistent with Section 3.2.1.2 and analyzed as described in Section 3.7.3.8.

Solid Waste Management System

2) Oily waste is part of the LRWS. Oily waste influent from the LRWS is packaged in 55-gallon drums in the RWB and then shipped offsite.

cale US460 SDAA 11.4-13 Revision 1

NuScale Final Safety Analysis Report Solid Waste Management System Figure 11.4-2a: Process Flow Diagram for Wet Solid Waste TO RWBVS TO SPENT RESIN STORAGE TANK FC TO PHASE SEPARATOR TANK FC FROM SERVICE AIR TO SPENT RESIN STORAGE TANK FC FROM SPENT RESIN STORAGE TANK FC TURB AC FROM LRWS PROCESSING SKIDS FC FC FC TO RWBVS FC AC TO RWBVS FC FQ DEWATERING SKID PACKAGE FC FC TO LRWS LCW COLLECTION TANKS SLOPE FC FC FROM SERVICE AIR PT PT SLOPE LT LT FC FC PHASE SEPARATOR TANK A PHASE SEPARATOR TANK B LT LT FC FC FC FC RWDS RWDS FROM LRWS DW BREAK TANK FC LC TO SPENT RESIN STORAGE TANK FC FC FC LC FC PIT FC DRAIN TO RWDS DRAIN TO RWDS FROM SERVICE AIR FC LC SPENT RESIN STORAGE TANK DRAIN TO RWDS TRANSFER PUMP B FC PIT TO LRWS PROCESSING SKIDS FC FE TO LCW COLLECTION TANKS FC FC LC SPENT RESIN STORAGE TANK TRANSFER PUMP A TO CVCS IX DRAIN TO RWDS FC NuScale US460 SDAA 11.4-14 Revision 1

NuScale Final Safety Analysis Report Solid Waste Management System Figure 11.4-2b: Solid Radioactive Waste System Diagram FROM SERVICE AIR SLOP E

TURB TO DEWATERING SKID AC FC FROM PCWS SPENT RESIN FC FC TO RWBVS FROM CVCS IX (SPENT RESIN) FC FC FC FC E FC SLOP FC FC E

PT SLOP LT PT FC LT FC FROM LRWS OIL SEPARATOR BBY1 COMPACTOR FC FC SPENT RESIN STORAGE TANK A SPENT RESIN STORAGE TANK B 55 GALLON LT LT DRUM FC FC FC FC RWDS RWDS FROM LRWS DW BREAK TANK FC LC FC FC FC LC DRAIN TO RWDS PIT FC DRAIN TO RWDS FC FC FC LC FROM SERVICE AIR SPENT RESIN STORAGE TANK DRAIN TO RWDS TRANSFER PUMP B FC PIT TO LCW COLLECTION TANKS FC FT TO PCWS DEMINERALIZERS FC FC LC SPENT RESIN STORAGE TANK TRANSFER PUMP A TO CVCS IX DRAIN TO RWDS FC NuScale US460 SDAA 11.4-15 Revision 1

The process and effluent radiological monitoring instrumentation and sampling design features provide the ability to detect and determine the content and, where required, the concentration and release rate of radioactive material in various gaseous and liquid process and effluent streams. The design features facilitate radiation monitoring and control, archiving, alarm functions and, where required, isolation and actuation functions to support the design objectives of the related system. The monitoring of in-plant radiation and airborne radioactivity is performed by the area radiation monitoring instrumentation described in Section 12.3.4.

5.1 System Description Effluent Radiation Monitoring is provided for:

  • Air cooled condenser system (ACCS) (Section 10.4.1)
  • Liquid radioactive waste system (LRWS) (Section 11.2)
  • Pool cooling and cleanup system (PCWS) (Section 9.1.3)
  • Reactor Building HVAC system (RBVS) (Section 9.4.2)
  • Site cooling water system (SCWS) (Section 9.2.7)
  • Utility water system (UWS) (Section 9.2.9)

Process Radiation Monitoring is provided for:

  • Auxiliary boiler system (ABS) (Section 10.4.7)
  • Balance-of-plant drain system (BPDS) (Section 9.3.3)
  • Chemical and volume control system (CVCS) (Section 9.3.4)
  • Condensate polisher resin regeneration system (CPS) (Section 10.4.5)
  • Containment evacuation system (CES) (Section 9.3.6)
  • Containment flooding and drain system (CFDS) (Section 9.3.7)
  • Normal control room HVAC system (CRVS) (Section 9.4.1)
  • Demineralized water system (DWS) (Section 9.2.3)
  • Gaseous radioactive waste system (GRWS) (Section 11.3)
  • Reactor component cooling water system (RCCWS) (Section 9.2.2)
  • Radioactive Waste Building HVAC system (RWBVS) (Section 9.4.3)
  • Radioactive waste drain system (RWDS) (Section 9.3.3)
  • Turbine generator system (TGS) (Section 10.2)

The following tables and figures provide a summary of radiological monitoring:

  • Detector information including number, type, location, and measurement range is provided in Table 11.5-1.

cale US460 SDAA 11.5-1 Revision 1

  • Effluent and process monitoring off-normal radiation conditions are described in Table 11.5-4.
  • Figure 11.5-1a and Figure 11.5-1b present an integrated plant radiological monitoring drawing.
  • Figure 11.5-2 provides a logic block diagram for radiation monitoring.
  • Figure 11.5-3 provides an off-line radiation detection drawing.
  • Figure 11.5-4 provides a process radiation adjacent-to-line detection drawing.
  • Figure 11.5-5 provides a process radiation in-line detection drawing.
  • Figure 11.5-6 provides a plant exhaust stack effluent radiation detection drawing.

Monitoring and operator response for effluent and process radiation monitors is performed in accordance with site procedures. Controls ensure that gaseous effluent content meet the objectives of 10 CFR 50 Appendix I and 10 CFR 20 before being released into the environment, and ensures compliance with GDC 60, 63, and 64.

Setpoints for radiation alarms (Section 11.5.1.2) and automated function initiation (Section 11.5.1.3) are based on ensuring that the limitations of 10 CFR 20 and 10 CFR 50 are met for plant conditions. Additionally, the alarms and isolations ensure compliance with GDC 60, 61, 63 and 64, and the applicable 10 CFR 20 and 10 CFR 50 requirements and limitations.

The ability to isolate and sample potentially contaminated systems ensures compliance to the occupation exposure limits in accordance with 10 CFR 20.1201 and 10 CFR 20.1202, and limits the spread of contamination per 10 CFR 20.1406.

Stack flow measurement capability supports the consideration of atmospheric dispersion (/Q) and deposition (D/Q) factors when developing alarm setpoints.

The RBVS plant exhaust stack flow rate and noble gas, particulate, and halogen activity indications are post-accident monitoring system variables as described in Table 7.1-7.

5.1.1 Reliability and Quality Assurance The quality assurance controls for digital computer software used in radiation monitoring and sampling equipment is described in Section 7.2.

Programs and procedures for the control of measuring and test equipment are administered per the quality assurance program described in Section 17.5.

5.1.2 Effluent Instrumentation Alarm Setpoints Effluent alarm setpoints are determined in accordance with the guidance of NUREG-1301 and NUREG-0133 such that effluent releases to unrestricted areas cale US460 SDAA 11.5-2 Revision 1

in the Offsite Dose Calculation Manual (ODCM), with consideration given to site-specific liquid effluent dilution factors and gaseous effluent atmospheric dispersion conditions.

5.1.3 Effluent Release Controls The gaseous and liquid effluent control for the plant is described in the ODCM and includes a description of how effluent release rates are derived and parameters used in setting instrumentation alarm setpoints to control or terminate effluent releases in unrestricted areas that are above the effluent concentrations in Table 2 of Appendix B to 10 CFR Part 20.

5.1.4 Offsite Dose Calculation Manual and Radiological Environmental Monitoring Program The ODCM contains a description of the methodology and parameters used for calculation of offsite doses for gaseous and liquid effluents. The ODCM also contains the planned effluent discharge flow rates and addresses the numerical requirements of 10 CFR 50, Appendix I.

The ODCM and Radiological Environmental Monitoring Program are developed and implemented in accordance with the recommendations and guidance of NEI 07-09A (Reference 11.5-2).

5.1.5 Process and Effluent Monitor Ranges The process and effluent radiation monitor instrument ranges are based on 10 CFR 20, Appendix B, and Regulatory Guides 1.21, 1.45, and 1.97.

5.2 References 11.5-1 American National Standards Institute/Health Physics Society, "Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities," ANSI/HPS N13.1-2011, Washington, DC.

11.5-2 Nuclear Energy Institute, "Generic FSAR Template Guidance for Offsite Dose Calculation Manual (ODCM) Program Description," NEI 07-09A, Revision 0, March 2009.

cale US460 SDAA 11.5-3 Revision 1

Scale Final Safety Analysis Report stem Quantity Type Service Isotopes Measurement Range Location/Function Safety- Media Instrument type related Cs-137 3E-10 to 1E-6 Ci/cc I-131 3E-10 to 5E-8 Ci/cc ABS Skid Vent to 1 ABS Skid Vent ATM Kr-85 3E-7 to 1E-2 Ci/cc No Gas Off-line (PING)

Atmosphere Xe-133 Ar-41 1E-7 to 1E-1 Ci/cc 1 ABS/Superheater skid Ar-41 1E-7 to 1E-1 Ci/cc ABS/Superheater Skid No Gas Adjacent-to-line BPDS Outlet BPDS Outlet Cs-137 3E-10 to 1E-6 Ci/cc I-131 3E-10 to 5E-8 Ci/cc ABS Skid to AB ABS/Superheater Skid 1 Kr-85 3E-7 to 1E-2 Ci/cc No Gas Off-line (PING)

Superheater Skid Vent BPDS Vent Xe-133 Ar-41 1E-7 to 1E-1 Ci/cc 1 Auxiliary Steam to BPD Ar-41 1E-7 to 1E-1 Ci/cc Auxiliary Steam to BPDS No Gas Adjacent-to-line 1 TGB Auxiliary Steam Ar-41 1E-7 to 1E-1 Ci/cc TGB Auxiliary Steam No Gas Adjacent-to-line Process and Effluent Radiation Monitoring Instrumentation and Header Header S 6 SJAE Gaseous Effluent Ar-41 1E-6 to 1E-1 Ci/cc SJAE Vent Air Evacuation No Gas Adjacent-to-line Line S 1 LRVP Gaseous Effluent Ar-41 1E-6 to 1E-1 Ci/cc LRVP Vent Air Evacuation No Gas Adjacent-to-line Line Cs-137 3E-10 to 1E-6 Ci/cc I-131 3E-10 to 5E-8 Ci/cc CARS Common Vent Air CARS Common Vent Air S 1 Kr-85 3E-7 to 1E-2 Ci/cc No Gas Off-line (PING)

Evacuation Line Evacuation Line Xe-133 Ar-41 1E-6 to 1E-1 Ci/cc Sampling System S 1 Turbine Building Floor Cs-137 1E-7 to 1E-2 µCi/ml Turbine Building Floor No Liquid In-line Drains Radiation Drains Radiation S 1 Condensate Cs-137 1E-7 to 1E-2 µCi/ml Condensate Regeneration No Liquid In-line Regeneration Skid Skid S 1 Aux Boiler Blowdown Cs-137 1E-7 to 1E-2 µCi/ml Aux Boiler Blowdown No Liquid In-line 6 Sample Tank Liquid Cs-137 1E-7 to 1E-1 µCi/ml Sample Tank Liquid No Liquid Adjacent-to-line Radiation Radiation

Scale Final Safety Analysis Report stem Quantity Type Service Isotopes Measurement Range Location/Function Safety- Media Instrument type related Cs-137 3E-10 to 1E-6 Ci /cc Gaseous Discharge Gaseous Discharge I-131 3E-10 to 5E-8 Ci /cc 6 Particulate, Noble Gas, Particulate, Noble Gas, No Gas Off-line (PING)

Iodine Radiation Kr-85 3E-7 to 3E-2 Ci /cc Iodine Radiation Xe-133 6 Gaseous Discharge Ar-41 1E-7 to 1E-1 Ci /cc Gaseous Discharge Ar-41 No Gas Off-line Ar-41 Radiation Radiation S 1 CFDS Drain Separator Kr-85 3E-7 to 1E-2 Ci /cc CFDS radiation flow to No Gas Off-line Gas Discharge Xe-133 HEPA filter 1 Resin Transfer Cs-137 1E-7 to 1E-2 Ci /ml Turbine Generator No Liquid Adjacent-to-line Radioisotope Building (TGB)/Condenser concentration Polisher Resin Regen Skid Inlet 1 Regeneration Sump Cs-137 1E-7 to 1E-2 Ci /ml TGB/Regeneration Sump No Liquid Adjacent-to-line Radioisotope Concentration Process and Effluent Radiation Monitoring Instrumentation and S 2 CRVS Outside Air Intake Kr-85 1E-5 to 1E+1 Rad/hr Outside Air Intake No Air Off-line Radiation Monitor Xe-133 Cs-137 3E-10 to 1E-4 Ci /cc CRE Supply Air I-131 1E-9 to 1E+2 Ci /cc S 2 CRE Supply Duct No Gas Off-line (PING)

Radiation Kr-85 4E-5 to 1E+4 Ci /cc Xe-133 S 6 RPV Supply to Cs-137 1E-7 to 1E-2 Ci /ml RPV Supply to RHX No Liquid Adjacent-to-line regenerative heat Cooling Inlet Radiation exchanger (RHX)

Cooling Inlet Radiation S 1 DW North Reactor Cs-137 1E-7 to 1E-2 Ci /ml DW North RXB No Liquid Off-line Sampling System Building (RXB) Contaminated Header Contaminated Header Radiation Radiation S 1 DW South RXB Cs-137 1E-7 to 1E-2 Ci /ml DW South RXB No Liquid Off-line Contaminated Header Contaminated Header Radiation Radiation WS 2 Charcoal Decay Bed Kr-85 3.0E-7 to 1.0E-2 Ci/ml GRW Charcoal Decay No Gas Off-line Skid A/B Outlet Xe-133 Bed Skid A/B Outlet Radiation Radiation

Scale Final Safety Analysis Report stem Quantity Type Service Isotopes Measurement Range Location/Function Safety- Media Instrument type related WS 1 GRW Outlet RadiationKr-85 3.0E-7 to 1.0E-2 Ci/ml GRW Outlet Radiation No Gas Off-line Xe-133 S 2 LRWS LCW Processing Cs-137 1.0E-7 to 1.0E-2 Ci/ml LWR discharge path No Liquid Adjacent to-line to UW radiation 6 SG1 Main Steam Line Ar-41 1.0E-7 to 1.0E-1 Ci/cc Main Steam Line No Gas Adjacent to-line Radiation 2 per steam header in the RXB 6 SG2 Main Steam Line Ar-41 1.0E-7 to 1.0E-1 Ci/cc Main Steam Line No Gas Adjacent to-line Radiation 2 per steam header in the RXB S 1 Surge Control Storage Kr-85 3.0E-7 to 1.0E-2 Ci/ml Monitor the PCWS surge No Gas Off-line Tank Vent Xe-133 control storage tank vent line on PCWS Cs-137 3E-10 to 1E-6 µCi/cc Spent Fuel Pool (SFP)

I-131 3E-10 to 5E-8 µCi/cc SFP Exhaust Filter S 1 Exhaust Filter Upstream No Gas Off-line (PING)

Process and Effluent Radiation Monitoring Instrumentation and Kr-85 3E-7 to 1E-2 µCi/cc Upstream Radiation Xe-133 Reactor Pool and Gallery Cs-137 3E-10 to 1E-6 µCi/cc Reactor Pool and Gallery S 1 No Gas Off-line (PI)

Area Exhaust I-131 3E-10 to 5E-8 µCi/cc Area Exhaust Cs-137 1E-7 to 1E-2 µCi/cc I-131 3E-10 to 1E-6 µCi/cc S 1 RBVS Exhaust Stack RBVS Exhaust Stack No Gas Off-line (PING)

Kr-85 3E-7 to 1E+4 µCi/cc Xe-133 RBVS South and North Cs-137 3E-10 to 1E-6 µCi/cc RBVS North and South Module Battery Rooms I-131 3E-10 to 5E-8 µCi/cc S 2 Module Battery Rooms No Gas Off-line (PING) air-handling unit (AHU) Kr-85 Sampling System 3E-7 to 1E-2 µCi/cc AHU Radiation Radiation Xe-133 Cs-137 3E-10 to 1E-6 µCi/cc SFP Exhaust Filter I-131 3E-10 to 5E-8 µCi/cc SFP Exhaust Filter S 1 No Gas Off-line (PING)

Upstream Radiation Kr-85 3E-7 to 1E-2 µCi/cc Upstream Xe-133

Scale Final Safety Analysis Report stem Quantity Type Service Isotopes Measurement Range Location/Function Safety- Media Instrument type related WS 6 RCCW CVCS Non- CS-137 1E-7 to 1E-2 µCi/ml Located in the RCCW- No Liquid Adjacent-to-line regenerative heat downstream of loads that exchanger Outlet have potential for a Radiation radioactive release to alert the control room when there is a leak in the RCCWS WS 6 RCCW CES Condenser CS-137 1E-7 to 1E-2 µCi/ml Located in the RCCWS No Liquid Adjacent-to-line and Vacuum Pumps downstream of loads that Outlet Radiation have potential for a radioactive release to alert the control room when there is a leak in the RCCWS WS 1 RCCW PSS Primary CS-137 1E-7 to 1E-2 µCi/ml Located in the RCCWS No Liquid Adjacent-to-line Sample Chiller Outlet downstream of all cooled Process and Effluent Radiation Monitoring Instrumentation and Radiation components RWBV Exhaust CS-137 3E-10 to 1E-6 µCi/cc Located in the RWBV Radiation Monitoring I-131 3E-10 to 5E-8 µCi/cc main exhaust duct before BVS 1 No Gas Off-line (PI)

Skid connecting to the RBVS exhaust duct DS 1 RXB RCCWS Drain Cs-137 1E-7 to 1E-2 Ci /ml RCCWS drain tank No Liquid Off-line Tank Radiation S 3 PCWS heat exchanger Cs-137 1E-7 to 1E-2 Ci /ml Downstream of RXB Heat No Liquid In-line A/B/C Outlet Radiation Exchangers/ PCWS outlet radiation S 1 CT Blowdown Radiation C-137 1E-7 to 1E-2 Ci /ml Cooling Tower Blowdown No Liquid Off-line with Sampling System Line Sampling S 2 RCCW Heat Exchanger Cs-137 1E-7 to 1E-2 Ci /ml Downstream of RXB Heat No Liquid In-line A/B Outlet Radiation Exchangers/ RCCW outlet radiation Cs-137 3E-10 to 1E-6 Ci/cc I-131 3E-10 to 5E-8 Ci/cc Turbine generator skid 6 Gland Steam Outlet Kr-85 3E-7 to 1E-2 Ci/cc common exhaust vent No Gas Off-line (PING)

Xe-133 point particulates Ar-41 1E-6 to 1E-1 µCi/cc

Scale Final Safety Analysis Report stem Quantity Type Service Isotopes Measurement Range Location/Function Safety- Media Instrument type related S 1 Letdown Line Radiation Cs-137 1E-7 to 1E-2 µCi/ml UWS effluent path No Liquid Off-line eral Notes:

Main control room monitoring available Waste management control room monitoring available Local Monitoring Available Designed to meet ANSI/HPS N13.1-2011 (Reference 11.5-1)

G - Particulate, Iodine, Noble Gas Particulate, Iodine Process and Effluent Radiation Monitoring Instrumentation and Sampling System

No. Gaseous Process or Waste System Sample Provisions(a)

Process Effluent Auxiliary Boiler System I, S&A Air Cooled Condensing System I, S&A S&A Containment Evacuation System I, S&A Control Room HVAC System I Containment Flooding Drain System S&A Gaseous Radioactive Waste System I Main Steam System S&A Pool Cooling & Clean Up System S&A S&A, H3 Radioactive Waste Building HVAC System I -

Reactor Building HVAC System I NG, H3 Reactor Building HVAC System (Spent Fuel Area) I Turbine Generator System I, S&A Sample point is available to obtain grab samples for laboratory analyses.

- Noble gas radioactivity Iodine radioactivity Tritium

-Sampling and analysis of radionuclides, including gross radioactivity, identification, and concentration of principal or significant radionuclides, and concentration of alpha emitters.

cale US460 SDAA 11.5-9 Revision 1

No. Liquid Process or Waste System Sample Provisions(a)

Process Effluent Balance of Plant Drain System S&A, H3 -

Containment Evacuation System(c) S&A -

Condensate Polisher Resin Regeneration System S&A -

Chemical and Volume Control System S&A, H3 -

Demineralized Water System S&A, H3 -

Liquid Radioactive Waste System(b) S&A S&A, H3 Reactor Component Coolant Water System S&A Radioactive Waste Drain System(b) S&A -

Site Cooling Water System S&A S&A, H3 Utility Water System S&A S&A,H3 Sample point is available to obtain grab samples for laboratory analyses.

The provisions for sampling potentially contaminated system and the use of the RWDS and LRWS for waste collection in the Reactor Building ensure compliance to occupational exposure limits in accordance with 10 CFR 20.1201 and 10 CFR 20.1202, and limit contamination per 10 CFR20.1406.

An installed mechanical liquid grab sampler located downstream of the CES sample vessel allows for the samples to be taken and analyzed in the laboratory for a more finite definition of the radionuclide content of the condensate, and to serve as a redundant means of measuring process radiation level. The mechanical sampler is designed to conform with RGs 8.8 and 8.10 and enhance plant staff capability to meet ALARA goals and contamination control in accordance with 10 CFR 20.1406. Compliance with RG 1.45 requirements and the capabilities of the CES sample vessel are discussed in Section 5.2.5.

-Noble gas radioactivity Iodine radioactivity Tritium

-Sampling and analysis of radionuclides, including gross radioactivity, identification and concentration of principal or significant radionuclides, and concentration of alpha emitters.

cale US460 SDAA 11.5-10 Revision 1

Conditions System Condition System Response Radiation Detected If high radiation is detected in the auxiliary boiler system skid vents, skid drains, or steam header drains, then the auxiliary boiler superheater skid outlet valve closes, auxiliary boiler skid to superheater skid valve closes, the module-specific main steam to auxiliary boiler header valves close, and the MCR receives an alarm.

S High Radiation Upon alarm, the wastewater collection tank pumps are shut down, the discharge isolation valves directing the disposition of the water are both closed, and manual intervention is initiated. The chemical waste collection tank is also monitored for radiation and upon radiation detection, the two affected chemical waste collection tank pumps are disabled and the two affected discharge isolation valves are closed. The radiation monitor alarms in the main control room for action and the RWBS control for information and an operator is dispatched to assess the situation.

High Radiation Upon detection, the Purge Gas Supply to the vacuum pumps are shut off and that discharge path is switched from the RBVS to the GRWS valve. The SA Connection valve receives a close signal.

Radiation Detection Spent resin being sent to the condensate polisher resin regeneration skid and the regeneration sump are monitored for radiation. If radiation is detected, a local alarm and an alarm in the MCR alerts operations staff.

S High Radiation levels continue Upon detection of a "high" radiation level in the to degrade outside air intake, the system is realigned so that 100 percent of the outside air passes through the or CRVS filter unit, containing HEPA and charcoal filters, to filter outside air and minimize radiation exposure to Radiation monitor power loss personnel with the CRB.

If power is not available to either CRVS AHU or to any of the four EDS-C battery chargers (after a 10-minute time delay), or if levels of radiation greater than 10 times background in the CRE supply air duct or if toxic gas is detected in the CRE supply air duct, the PPS automatically isolates the CRE from the adjacent areas by closing the redundant CRE isolation dampers). The time delay is to allow operators time to restore power and start the stand-by AHU. The operating supply AHU and associated components, the general exhaust fan, and the battery exhaust fan are also turned off and the CRH is automatically initiated. The CRH provides a supply of breathable air for the CRE occupants and maintains the CRE at a positive pressure with respect to the surrounding areas. The heat sink capacity of surrounding structures of the CRE helps maintain the temperature in the CRE within acceptable tolerances.

cale US460 SDAA 11.5-11 Revision 1

System Condition System Response S High Radiation Process sample line isolation or Radiation monitor power loss S Radiation Alarm PCS alarms and automatically close the associated upstream on-off valve.

WS High radiation level in a decay Close the inlet valve and outlet valve from the affected bed skid outlet decay bed skid as well as the outlet valves to RWBVS.

WS High radiation level in the Close the outlet valves to the RWBVS to stop the connection line to the RWBVS system flow.

WS High radiation level in charcoal Close the inlet valves to the GRWS to stop the system bed cubicle flow. Open the nitrogen purge valve.

S High Radiation on Sample Pump to Collection Tank for reprocessing (manual Tank operator action)

S High Radiation on single Point Table 11.2-2 LRW discharge High Radiation If a high radiation condition is detected on the main steam line radiation monitors, an alarm in the MCR cues the operators to take actions to mitigate the event per applicable operating procedures. If required, the main steam lines can be manually isolated from the MCR. Additionally, the MSS drain pots automatically isolate during high radiation for both normal operation or when a MPS isolation signal is present in order to ensure that the MSS does not contribute to unmonitored release of high radioactivity to the environment in the event of an abnormal tube leak. High radiation detection provides an alarm in the control room. If the drain pots and/or isolation valves to the ACC CCT are open, they close. Operator action from the MCR can isolate the MS lines, if required.

S High Radiation Alarm in MCR to initiate appropriate safety actions.

S High Radiation Alarm in MCR but no automatic actions. Operating staff takes actions to determine the source of the contamination and isolate it.

S (SFP Area) High Radiation SFP exhaust is diverted through both the HEPA filters and the charcoal absorbers. Isolation dampers of the RXB general exhaust fans reduce speed in response to the damper closures to maintain the design exhaust header setpoint.

BVS High Radiation Upon high limit detection of radiation in the RWBVS exhaust effluent to the RBVS system, action is taken by plant operators to locate the source of contamination, however RWBVS continues to operate.

DS High Radiation PCS alarms and interlock closes the valve back to the RCCW expansion tank.

cale US460 SDAA 11.5-12 Revision 1

System Condition System Response S Radiation Detection Upon alarm, operators are alerted to abnormal condition, prompting them to investigate and isolate leaks or terminate other conditions that contribute to the off-normal conditions, through valve closures.

S High Radiation Alarm in the MCR and locally cale US460 SDAA 11.5-13 Revision 1

Scale Final Safety Analysis Report Radioactive Effluent Flow Paths With Process and Effluent Radiation Monitors (Inside Reactor Building)

R From Pool Cooling & Cleanup System Containment To Reactor Building Ventilation System Flood and Drain System To Pool Cooling & Cleanup System R

To Turbine Generator System Main Steam R

System R

R To Gaseous Radwaste System Containment To Radioactive Waste Drain System Evacuation To Reactor Building Ventilation System System R

To Liquid Radwaste System Reactor Process Component Sampling Cooling Water System Process and Effluent Radiation Monitoring Instrumentation and To Air Cooled Condenser System R

R Chemical and Volume Control To Liquid Radwaste System System Reactor R Building To Reactor Building Ventilation System Ventilation Intake Containment Vessel Sampling System RBVS SFP To Reactor Building Ventilation System Exhaust R

Scale Final Safety Analysis Report Radioactive Effluent Flow Paths With Process and Effluent Radiation Monitors (Outside Reactor Building)

R R From RBVS SPF Exhaust R

R Reactor From Containment Flood & Drain System Building From Process Sampling System From Containment Evacuation System Ventilation Auxiliary Boiler Air Cooled System System Condenser From Reactor Building Ventilation Intake R

System R R

Radwaste Condensate Gaseous Building Polisher Resin From Containment Evacuation System Radwaste Ventilation Regeneration System System System From Containment Evacuation System R Radioactive Control Room Waste Drain Ventilation System System From Chemical & Volume Control System Liquid From Process Sampling System Radwaste System Process and Effluent Radiation Monitoring Instrumentation and R

From Containment Flood & Drain System Pool Cooling and Cleanup To Containment Flood & Drain System R

System Balance of Plant Demineralized Site Cooling Drain System Water System Water System R

R From Main Steam System R Turbine Generator R Sampling System System Utility Water System Site Drainage System

Scale Final Safety Analysis Report Configuration Radiation Monitor Radiation Detector Local Signal Conditioning / MPS / PPS Protective Functions Electronic Control Unit Safety Display &

Indication System Display Functions Process and Effluent Radiation Monitoring Instrumentation and MCS / PCS Display, MCS / PCS Alarm and Data Recording Functions Radiation Monitor MCS / PCS Control, Sampling System Local Signal Conditioning / Display, Alarm and Radiation Detector Electronic Control Unit MCS / PCS Data Recording Functions

Scale Final Safety Analysis Report (Indication Note 1) (Note1:

MCS, PCS, PPS Indication may be provided to either or all I&C systems listed)

Local Signal Conditioning /

Electronic Control Unit Inlet Isolation Valve Outlet Isolation Valve RE Process and Effluent Radiation Monitoring Instrumentation and Chamber Sample Pump Process Line Sampling System Off Line Radiation Monitor

Scale Final Safety Analysis Report (Indication Note 1) (Note1:

MCS, PCS Indication may be provided to either or all I&C systems listed)

Local Signal Conditioning /

Electronic Control Unit RE Process and Effluent Radiation Monitoring Instrumentation and Process Line Adjacent to Line Radiation Monitor Sampling System

Scale Final Safety Analysis Report (Indication Note 1)

MCS, PCS (Note1: Indication may be provided to either or all I&C systems listed)

Local Signal Conditioning /

Electronic Control Unit Process and Effluent Radiation Monitoring Instrumentation and RE Process Line In Line Radiation Monitor Sampling System

Scale Final Safety Analysis Report PCS FE FT Signal Conditioning Processing Unit RE RE RE Isokinetic Sample Array Particulate Iodine Gas Process and Effluent Radiation Monitoring Instrumentation and Normal Range Accident Range RBVS PLANT VENT RE RE Particulate Iodine Gas (Mid) Gas (High)

Sampling System

Section 11.5 discusses effluent and process radiation monitors. This discussion contains the radiation monitoring (RM) design functions, features, and bases for the plant systems containing effluent or process radiation monitors and includes a discussion of the compliance with associated regulatory requirements and guidance documents.

Section 11.5 discusses provisions for sampling in the systems containing effluent and process radiation monitors. For selected systems, these provisions include functions provided by the process sampling system, which is discussed in Section 9.3.2.

Section 12.3.4 discusses area radiation and airborne contamination monitors. This discussion contains the RM design functions, features, and bases for plant area radiation and airborne contamination monitors and includes a discussion of the compliance with associated regulatory requirements and guidance documents.

Section 7.2.13 describes effluent and area RM that provide input to the emergency response data system and the electronic data communication interface. Section 11.5 and Section 12.3.4 address effluent and area radiological monitoring parameters and equipment.

Applicable portions of the quality assurance program described in Section 17.5 administer the programs and procedures for the control of measuring and test equipment.

Section 17.6 describes the program for monitoring the effectiveness of maintenance.

cale US460 SDAA 11.6-1 Revision 1

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